2020
D Gallart; M J Mantsinen; P Jacquet; K Kirov; E. Lerche; J Wright; JET Contributors
Modelling of ICRF heating for JET T and D-T plasmas Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 060001, 2020.
@article{1981,
title = {Modelling of ICRF heating for JET T and D-T plasmas},
author = {D Gallart and M J Mantsinen and P Jacquet and K Kirov and E. Lerche and J Wright and JET Contributors},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {060001},
abstract = {A tritium (T) campaign is planned in preparation for the deuterium-tritium (D-T) campaign at the Joint European Torus (JET). These experiments will be the first experiments involving T with the ITER-like plasma-wall facing components materials. They will give a unique opportunity to test one of the most promising ion cyclotron resonance frequency (ICRF) heating schemes for ITER plasmas: the 2nd tritium (T) harmonic resonance (ω = 2ωT ). This paper provides two key contributions related to modelling of the performance of this scheme at JET. First, we assess the heating performance of the 2nd T harmonic resonance and, second, we model different ICRF schemes for the T campaign in support for the D-T campaign, i.e., identify differences and similaritiesfromtheheatingpointofviewbetweenTandD-Tplasmasinordertopredicttheperformanceofω=2ωT intheD-T scenario. In our modelling we use a selected hybrid record discharge as reference, i.e., using its experimental profiles. We consider two ICRF schemes (6 MW), i.e. ω = ω3He = 2ωT (no 3He), with a central resonance and three NBI power outputs (15, 25 and 35 MW) in two plasma compositions ( 100% T and 50%:50% D-T). Note that isotope effects are not taken into account in these simulations. For this study, the ICRF and NBI heating are modelled with the ICRF code PION and the beam code PENCIL which take into account the ICRF+NBI synergy. The analysis of the T velocity distribution function shows that a stronger tail is formed in those plasmas with lower tritium density. This fact has an important impact on the slowing-down process of fast tritons with the background species. The T plasma shows a higher and more peaked ion-ion collisional power density at the plasma centre as compared to D-T plasma. In the T plasma, ICRF heating drives fast tritons at the plasma centre with an average energy substantially lower than in the other case. On the other hand, the use of 3He as a minority makes the fast ion T energy considerably lower due to strong 3He absorption. Fast ion average energies reached at the plasma centre are similar in all species mixture cases. As a result, there is a strong heating similarity between T and D-T. However, it is crucial to study the generation of a strong T tail as a result of particle-wave interaction which is not possible under this scheme and needs to be studied in plasmas without 3He in preparation of JET D-T campaign and ITER. Therefore, both schemes (ω = ω3 He = 2ωT and ω = 2ωT ) need to be tested.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
A tritium (T) campaign is planned in preparation for the deuterium-tritium (D-T) campaign at the Joint European Torus (JET). These experiments will be the first experiments involving T with the ITER-like plasma-wall facing components materials. They will give a unique opportunity to test one of the most promising ion cyclotron resonance frequency (ICRF) heating schemes for ITER plasmas: the 2nd tritium (T) harmonic resonance (ω = 2ωT ). This paper provides two key contributions related to modelling of the performance of this scheme at JET. First, we assess the heating performance of the 2nd T harmonic resonance and, second, we model different ICRF schemes for the T campaign in support for the D-T campaign, i.e., identify differences and similaritiesfromtheheatingpointofviewbetweenTandD-Tplasmasinordertopredicttheperformanceofω=2ωT intheD-T scenario. In our modelling we use a selected hybrid record discharge as reference, i.e., using its experimental profiles. We consider two ICRF schemes (6 MW), i.e. ω = ω3He = 2ωT (no 3He), with a central resonance and three NBI power outputs (15, 25 and 35 MW) in two plasma compositions ( 100% T and 50%:50% D-T). Note that isotope effects are not taken into account in these simulations. For this study, the ICRF and NBI heating are modelled with the ICRF code PION and the beam code PENCIL which take into account the ICRF+NBI synergy. The analysis of the T velocity distribution function shows that a stronger tail is formed in those plasmas with lower tritium density. This fact has an important impact on the slowing-down process of fast tritons with the background species. The T plasma shows a higher and more peaked ion-ion collisional power density at the plasma centre as compared to D-T plasma. In the T plasma, ICRF heating drives fast tritons at the plasma centre with an average energy substantially lower than in the other case. On the other hand, the use of 3He as a minority makes the fast ion T energy considerably lower due to strong 3He absorption. Fast ion average energies reached at the plasma centre are similar in all species mixture cases. As a result, there is a strong heating similarity between T and D-T. However, it is crucial to study the generation of a strong T tail as a result of particle-wave interaction which is not possible under this scheme and needs to be studied in plasmas without 3He in preparation of JET D-T campaign and ITER. Therefore, both schemes (ω = ω3 He = 2ωT and ω = 2ωT ) need to be tested.
W Helou; F. Durodié; J Hillairet; E. Lerche; G Lombard; D Milanesio; P Mollard; J -M Bernard; N Charabot; F Clairet; L Colas; Y Moudden; B Santraine; Y T Song; G Urbanczyk; R Volpe; Y S Wang; Q X Yang; V Bobkov; WEST Team
Characterizations and first plasma operation of the WEST load-resilient actively cooled ICRF launchers Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 030009, 2020.
@article{1972,
title = {Characterizations and first plasma operation of the WEST load-resilient actively cooled ICRF launchers},
author = {W Helou and F. Durodié and J Hillairet and E. Lerche and G Lombard and D Milanesio and P Mollard and J -M Bernard and N Charabot and F Clairet and L Colas and Y Moudden and B Santraine and Y T Song and G Urbanczyk and R Volpe and Y S Wang and Q X Yang and V Bobkov and WEST Team},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {030009},
abstract = {The paper discusses the characterization of the three high power steady-state and load-resilient ICRF launchers of WEST before their installation in the tokamak. These launchers have been characterized and validated in low-power experiments (milliwatt range) as well as in experiments at the nominal RF voltages and currents in the TITAN vacuum chamber (~30 kV and 915 A peak). The successful commissioning of two of the launchers during the WEST C3 campaign at ~1 MW power level is illustrated. Manual and real-time controlled impedance-matching of the launchers are discussed, as well as the validation of their load-resilience. Furthermore, several redundant and complementary protection systems have been validated and are reviewed in the paper.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The paper discusses the characterization of the three high power steady-state and load-resilient ICRF launchers of WEST before their installation in the tokamak. These launchers have been characterized and validated in low-power experiments (milliwatt range) as well as in experiments at the nominal RF voltages and currents in the TITAN vacuum chamber (~30 kV and 915 A peak). The successful commissioning of two of the launchers during the WEST C3 campaign at ~1 MW power level is illustrated. Manual and real-time controlled impedance-matching of the launchers are discussed, as well as the validation of their load-resilience. Furthermore, several redundant and complementary protection systems have been validated and are reviewed in the paper.
K K Kirov; Y. O. Kazakov; M Nocente; J. Ongena; Y Baranov; F Casson; J Eriksson; L Giacomelli; C Hellesen; V Kiptily; R Bilato; K. Crombé; R Dumont; P Jacquet; T Johnson; E Lerch; M Mantsinen; D. Van Eester; J Varje; H Weisen; JET Contributors
Synergistic ICRH and NBI heating for fast ion generation and maximising fusion rate in mixed plasmas at JET Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 030011, 2020.
@article{1973,
title = {Synergistic ICRH and NBI heating for fast ion generation and maximising fusion rate in mixed plasmas at JET},
author = {K K Kirov and Y. O. Kazakov and M Nocente and J. Ongena and Y Baranov and F Casson and J Eriksson and L Giacomelli and C Hellesen and V Kiptily and R Bilato and K. Crombé and R Dumont and P Jacquet and T Johnson and E Lerch and M Mantsinen and D. Van Eester and J Varje and H Weisen and JET Contributors},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {030011},
abstract = {The studies of recent JET experiments in H/D≈0.85/0.15 plasma (2.9T/2MA) in which neutron rate was enhanced by applying 2.5MW of ICRH using D-(DNBI)-H three-ion scheme are reported. An extensive analysis of this novel heating scenario has been carried out by means of integrated TRANSP/TORIC modelling, and a comprehensive validation of the computed Fast Ion Distribution Function (FI DF) with a range of fast ion diagnostics available at JET is presented. The predicted acceleration of D Neutral Beam Injection (NBI) ions beyond their injection energies and the associated changes in FI DF by RF waves are found to be in good agreement with measured neutron yield and TOFOR neutron spectrometer measurements, as well as with multi-channel neutron camera observations and neutral particle analyser diagnostic. An outlook of the possible applications of the developed technique for future DTE2 studies on JET has been highlighted. Controlled acceleration of TNBI ions in D-rich and DNBI ions in T-rich plasmas to optimal energies can be applied to maximise BT fusion rates and contribute to the success of future DT experiments at JET and ITER as illustrated in this study.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The studies of recent JET experiments in H/D≈0.85/0.15 plasma (2.9T/2MA) in which neutron rate was enhanced by applying 2.5MW of ICRH using D-(DNBI)-H three-ion scheme are reported. An extensive analysis of this novel heating scenario has been carried out by means of integrated TRANSP/TORIC modelling, and a comprehensive validation of the computed Fast Ion Distribution Function (FI DF) with a range of fast ion diagnostics available at JET is presented. The predicted acceleration of D Neutral Beam Injection (NBI) ions beyond their injection energies and the associated changes in FI DF by RF waves are found to be in good agreement with measured neutron yield and TOFOR neutron spectrometer measurements, as well as with multi-channel neutron camera observations and neutral particle analyser diagnostic. An outlook of the possible applications of the developed technique for future DTE2 studies on JET has been highlighted. Controlled acceleration of TNBI ions in D-rich and DNBI ions in T-rich plasmas to optimal energies can be applied to maximise BT fusion rates and contribute to the success of future DT experiments at JET and ITER as illustrated in this study.
A Kostic; K. Crombé; R Dux; M Griener; M Usoltceva; I. Shesterikov; E H Martin; J. -M. Noterdaeme
Direct local electric field measurements in the sheaths of the ICRF antenna in IShTAR Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 050002, 2020.
@article{1978,
title = {Direct local electric field measurements in the sheaths of the ICRF antenna in IShTAR},
author = {A Kostic and K. Crombé and R Dux and M Griener and M Usoltceva and I. Shesterikov and E H Martin and J. -M. Noterdaeme},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {050002},
abstract = {An important step in understanding the Radio Frequency sheaths in the vicinity of Ion Cyclotron antennas in magnetically confined fusion plasmas is to benchmark the existing theoretical models against real antenna operation. The data needed is the local DC electric field that develops in the sheaths surrounding the plasma-facing structures of the antenna. In this contribution we present our most recent progress on the development of a spectroscopic diagnostic, which delivers direct and local measurements of the electric field across the sheath. The method uses polarization spectroscopy where the electric fields are optically determined from the Stark shift of the emission spectra of excited helium atoms. The sheath electric fields obtained with this diagnostic scheme are measured in the vicinity of the ICRF antenna in IShTAR, a test facility designed to mimic the tokamak edge plasma parameters and focused on studying the ICRF antenna-plasma interaction.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
An important step in understanding the Radio Frequency sheaths in the vicinity of Ion Cyclotron antennas in magnetically confined fusion plasmas is to benchmark the existing theoretical models against real antenna operation. The data needed is the local DC electric field that develops in the sheaths surrounding the plasma-facing structures of the antenna. In this contribution we present our most recent progress on the development of a spectroscopic diagnostic, which delivers direct and local measurements of the electric field across the sheath. The method uses polarization spectroscopy where the electric fields are optically determined from the Stark shift of the emission spectra of excited helium atoms. The sheath electric fields obtained with this diagnostic scheme are measured in the vicinity of the ICRF antenna in IShTAR, a test facility designed to mimic the tokamak edge plasma parameters and focused on studying the ICRF antenna-plasma interaction.
E. Lerche; D. Van Eester; P Jacquet; F Casson; Y Baranov; P. Dumortier; D Gallart; J Graves; P Huynh; T Johnson; Y. O. Kazakov; V Kiptily; K Kirov; M Machielsen; M Mantsinen; I Monakhov; J. Ongena; JET Contributors
ICRH options for JET-ILW DTE2 operation Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 030007, 2020.
@article{1971,
title = {ICRH options for JET-ILW DTE2 operation},
author = {E. Lerche and D. Van Eester and P Jacquet and F Casson and Y Baranov and P. Dumortier and D Gallart and J Graves and P Huynh and T Johnson and Y. O. Kazakov and V Kiptily and K Kirov and M Machielsen and M Mantsinen and I Monakhov and J. Ongena and JET Contributors},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {030007},
abstract = {Ion cyclotron resonance heating (ICRH) plays an important role in high performance JET-ILW plasma discharges, both for keeping the heavy impurities away from the plasma centre as for increasing the core ion temperature to boost fusion performance. While the former is needed in all high-performance discharges for steady state operation, the latter will be particularly important in the next-coming JET Deuterium-Tritium campaign (DTE2). Currently, the workhorse for impurity control in high power D plasmas is fundamental H minority ICRH (with simultaneous ω=2ωc harmonic D heating), which leads to localized core electron heating that induces turbulence (flatter density profiles) as well as peaked electron temperatures. For fusion power enhancement, dominant bulk ion heating and RF acceleration of the NBI ions to appropriate energies would be preferable and theoretical predictions suggest that ion heating is also effective for core impurity screening. In this paper, we discuss the basic modeling results of different ICRF scenarios available for the DTE2 campaign in JET-ILW, highlighting their main properties in terms of the RF absorption of the various species, their slowing-down properties and their impact on high-Z impurity transport. Correctly modeling the wave absorption, slowing-down and collisional energy redistribution of the simultaneously RF-heated species in a DT plasma mix with important neutral-beam injection (NBI) is numerically challenging and is outside the scope of this paper. The simplified calculations presented here are rather intended to give the reader an overview of the ICRH options for JET-DTE2 with references to the state-of-the-art ICRH modeling given throughout the paper.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Ion cyclotron resonance heating (ICRH) plays an important role in high performance JET-ILW plasma discharges, both for keeping the heavy impurities away from the plasma centre as for increasing the core ion temperature to boost fusion performance. While the former is needed in all high-performance discharges for steady state operation, the latter will be particularly important in the next-coming JET Deuterium-Tritium campaign (DTE2). Currently, the workhorse for impurity control in high power D plasmas is fundamental H minority ICRH (with simultaneous ω=2ωc harmonic D heating), which leads to localized core electron heating that induces turbulence (flatter density profiles) as well as peaked electron temperatures. For fusion power enhancement, dominant bulk ion heating and RF acceleration of the NBI ions to appropriate energies would be preferable and theoretical predictions suggest that ion heating is also effective for core impurity screening. In this paper, we discuss the basic modeling results of different ICRF scenarios available for the DTE2 campaign in JET-ILW, highlighting their main properties in terms of the RF absorption of the various species, their slowing-down properties and their impact on high-Z impurity transport. Correctly modeling the wave absorption, slowing-down and collisional energy redistribution of the simultaneously RF-heated species in a DT plasma mix with important neutral-beam injection (NBI) is numerically challenging and is outside the scope of this paper. The simplified calculations presented here are rather intended to give the reader an overview of the ICRH options for JET-DTE2 with references to the state-of-the-art ICRH modeling given throughout the paper.
F. Louche; F. Durodié; W Helou; P. U. Lamalle; F Calarco
Three-dimensional RF and circuit modelling of the revised ITER ICRF launcher design Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 070008, 2020.
@article{1986,
title = {Three-dimensional RF and circuit modelling of the revised ITER ICRF launcher design},
author = {F. Louche and F. Durodié and W Helou and P. U. Lamalle and F Calarco},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {070008},
abstract = {The ITER Ion Cyclotron Heating and Current Drive antenna design has undergone significant improvements driven by mechanical considerations. We present a series of numerical assessments of the RF properties of the resulting new architecture. Various designs aiming at using the service stub as a mechanical support were analysed and compared to the performance of the 2012 Preliminary Design Review (PDR) version taken as reference resulting in the selection of the ''internal service T-stub'' design. CST MWS/Ansys HFSS 3D models of the sub-components of the antenna were developed and simulated, and the resulting scattering matrices were integrated into transmission line (TL) models to evaluate RF coupling properties and RF power losses on the conductors. The results are compared with the 2012 performances and the differences are discussed. Also, 3D RF fields maps are produced and areas of improvement are identified.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The ITER Ion Cyclotron Heating and Current Drive antenna design has undergone significant improvements driven by mechanical considerations. We present a series of numerical assessments of the RF properties of the resulting new architecture. Various designs aiming at using the service stub as a mechanical support were analysed and compared to the performance of the 2012 Preliminary Design Review (PDR) version taken as reference resulting in the selection of the ''internal service T-stub'' design. CST MWS/Ansys HFSS 3D models of the sub-components of the antenna were developed and simulated, and the resulting scattering matrices were integrated into transmission line (TL) models to evaluate RF coupling properties and RF power losses on the conductors. The results are compared with the 2012 performances and the differences are discussed. Also, 3D RF fields maps are produced and areas of improvement are identified.
V. Maquet; A. Messiaen; R. Ragona
Study of the ohmic losses of a traveling wave antenna section in view of application on DEMO Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 070001, 2020.
@article{1983,
title = {Study of the ohmic losses of a traveling wave antenna section in view of application on DEMO},
author = {V. Maquet and A. Messiaen and R. Ragona},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {070001},
abstract = {A multi-section Traveling Wave Antenna (TWA) is one of the presently proposed options for Ion Cyclotron Resonance Heating (ICRH) of DEMO. This option would reduce the antenna power density while enabling sufficient heating in low coupling situations by increasing the number of radiating straps. Such an antenna consists of several sections each fed in a resonant ring con- figuration. In order to guide future engineering requirements, a detailed computation of the ohmic losses due to finite conductivity of the antenna components of one TWA section is performed for different loading conditions.
The geometry of the TWA section is compatible with the presently proposed layout of DEMO and is modeled by the finite element code Microwave Studio (MWS) taking into account the finite resistivity of its constituent materials. As MWS cannot handle plasma loading, it has been implemented in the code by a layer in front of the antenna with the real part of the permittivity of water and open boundary conditions to reproduce single-pass absorption conditions and capture the main antenna loading characteristics of a plasma. The loading conditions are then varied changing the distance between the antenna and this layer. An equivalent antenna resistance characterizing those ohmic losses as a function of frequency, material resistivity and loading has been computed. This resistance can be linked to the skin depth resistance and compared to the radiation resistance in order to assess the performance of the antenna for a given choice of material. These results are then extrapolated to a real plasma loading based on ANTITER results for different plasma profiles and water dielectric load distance.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
A multi-section Traveling Wave Antenna (TWA) is one of the presently proposed options for Ion Cyclotron Resonance Heating (ICRH) of DEMO. This option would reduce the antenna power density while enabling sufficient heating in low coupling situations by increasing the number of radiating straps. Such an antenna consists of several sections each fed in a resonant ring con- figuration. In order to guide future engineering requirements, a detailed computation of the ohmic losses due to finite conductivity of the antenna components of one TWA section is performed for different loading conditions.
The geometry of the TWA section is compatible with the presently proposed layout of DEMO and is modeled by the finite element code Microwave Studio (MWS) taking into account the finite resistivity of its constituent materials. As MWS cannot handle plasma loading, it has been implemented in the code by a layer in front of the antenna with the real part of the permittivity of water and open boundary conditions to reproduce single-pass absorption conditions and capture the main antenna loading characteristics of a plasma. The loading conditions are then varied changing the distance between the antenna and this layer. An equivalent antenna resistance characterizing those ohmic losses as a function of frequency, material resistivity and loading has been computed. This resistance can be linked to the skin depth resistance and compared to the radiation resistance in order to assess the performance of the antenna for a given choice of material. These results are then extrapolated to a real plasma loading based on ANTITER results for different plasma profiles and water dielectric load distance.
The geometry of the TWA section is compatible with the presently proposed layout of DEMO and is modeled by the finite element code Microwave Studio (MWS) taking into account the finite resistivity of its constituent materials. As MWS cannot handle plasma loading, it has been implemented in the code by a layer in front of the antenna with the real part of the permittivity of water and open boundary conditions to reproduce single-pass absorption conditions and capture the main antenna loading characteristics of a plasma. The loading conditions are then varied changing the distance between the antenna and this layer. An equivalent antenna resistance characterizing those ohmic losses as a function of frequency, material resistivity and loading has been computed. This resistance can be linked to the skin depth resistance and compared to the radiation resistance in order to assess the performance of the antenna for a given choice of material. These results are then extrapolated to a real plasma loading based on ANTITER results for different plasma profiles and water dielectric load distance.
A V Melnikov; J. Ongena; A. Messiaen; R. Ragona; A V Sushkov; Y. O. Kazakov; D. Van Eester; Yu. N Dnestrovskii; P P Khvostenko; I N Roy
Conceptual study of an ICRH traveling wave antenna (TWA) for T-15MD at 60 MHz Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 070007, 2020.
@article{1985,
title = {Conceptual study of an ICRH traveling wave antenna (TWA) for T-15MD at 60 MHz},
author = {A V Melnikov and J. Ongena and A. Messiaen and R. Ragona and A V Sushkov and Y. O. Kazakov and D. Van Eester and Yu. N Dnestrovskii and P P Khvostenko and I N Roy},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {070007},
abstract = {This paper proposes a Traveling Wave Antenna for ICRH heating of T-15MD located below the equatorial plane of the tokamak for operation at 60MHz. Resonant ring feeding allows the recirculation of the RF power that is not radiated to the plasma and the termination of the TWA section in its iterative impedance. The paper describes the antenna design, the feeding ring tuning algorithm and expected performances of this antenna concept. The chosen geometry of the TWA sections is compatible with that of a future reactor and therefore this ICRH antenna system for T-15MD also represents a test bed for DEMO.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
This paper proposes a Traveling Wave Antenna for ICRH heating of T-15MD located below the equatorial plane of the tokamak for operation at 60MHz. Resonant ring feeding allows the recirculation of the RF power that is not radiated to the plasma and the termination of the TWA section in its iterative impedance. The paper describes the antenna design, the feeding ring tuning algorithm and expected performances of this antenna concept. The chosen geometry of the TWA sections is compatible with that of a future reactor and therefore this ICRH antenna system for T-15MD also represents a test bed for DEMO.
A. Messiaen; R. Ragona; R. Koch; V. Maquet; J. Ongena
Effect of poloidal magnetic field and cross-coupling on a set of traveling wave antenna sections for the ICRH of fusion reactor plasmas Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 060005, 2020.
@article{1982,
title = {Effect of poloidal magnetic field and cross-coupling on a set of traveling wave antenna sections for the ICRH of fusion reactor plasmas},
author = {A. Messiaen and R. Ragona and R. Koch and V. Maquet and J. Ongena},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {060005},
abstract = {A set of Traveling Wave antenna (TWA) sections has been proposed for the ICRF heating of the reactor in view to decrease the antenna power density. The reduction of the antenna voltage and associate electric field and current is compensated by the number of radiating straps. An upgraded version of the fast semi-analytical code ANTITER-II is used to model a set of TWA sections of any arbitrary number of radiating straps facing a low coupling plasma profile with their feeding system. Now it incorporates the effect of the non-alignment of the antenna to the total steady magnetic field (toroidal + poloidal) in front of it. The model incorporates the feeding of each section by a resonant ring circuit that recirculates its output power. The cases of straps grounded at one of their ends (L grounding) or in their center (T grounding) are also compared. The model is applied to the proposed TWA section layout for the ICRF heating of DEMO. It is shown that the antenna tilting affects mostly the poloidal radiating spectrum seen by the plasma. This effect increases with the absolute value of the k// selected by the antenna system. The coupling reduction and the effect on the strap current and voltage distribution due to the tilting is compared with the ones resulting from the mutual coupling between the sections. The effects of the tilting for the expected qedge value of the reactor and of mutual coupling between toroidally spaced sections remain weak. The effect of coupling between poloidally superposed sections can become large.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
A set of Traveling Wave antenna (TWA) sections has been proposed for the ICRF heating of the reactor in view to decrease the antenna power density. The reduction of the antenna voltage and associate electric field and current is compensated by the number of radiating straps. An upgraded version of the fast semi-analytical code ANTITER-II is used to model a set of TWA sections of any arbitrary number of radiating straps facing a low coupling plasma profile with their feeding system. Now it incorporates the effect of the non-alignment of the antenna to the total steady magnetic field (toroidal + poloidal) in front of it. The model incorporates the feeding of each section by a resonant ring circuit that recirculates its output power. The cases of straps grounded at one of their ends (L grounding) or in their center (T grounding) are also compared. The model is applied to the proposed TWA section layout for the ICRF heating of DEMO. It is shown that the antenna tilting affects mostly the poloidal radiating spectrum seen by the plasma. This effect increases with the absolute value of the k// selected by the antenna system. The coupling reduction and the effect on the strap current and voltage distribution due to the tilting is compared with the ones resulting from the mutual coupling between the sections. The effects of the tilting for the expected qedge value of the reactor and of mutual coupling between toroidally spaced sections remain weak. The effect of coupling between poloidally superposed sections can become large.
J. Ongena; A. Messiaen; Y. O. Kazakov; B Schweer; I. Stepanov; M. Vervier; K. Crombé; M. Van Schoor; V Borsuk; Casta D n; A Kraemer-Flecken; K P Hollfeld; G Offermanns; D A Hartmann; J P Kallmeyer; R C Wolf
The ICRH system for the stellarator Wendelstein 7-X Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 070003, 2020.
@article{1984,
title = {The ICRH system for the stellarator Wendelstein 7-X},
author = {J. Ongena and A. Messiaen and Y. O. Kazakov and B Schweer and I. Stepanov and M. Vervier and K. Crombé and M. Van Schoor and V Borsuk and Casta D n and A Kraemer-Flecken and K P Hollfeld and G Offermanns and D A Hartmann and J P Kallmeyer and R C Wolf},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {070003},
abstract = {An important test for W7-X is to demonstrate confinement of fast trapped ions at volume averaged beta values up to 5%, corresponding to plasma densities above 1020 m-3. Energetic ions in W7-X with energies 50 < E < 100 keV mimic alphas in a reactor. To generate such a population is a challenging task in high-density plasmas but this can be efficiently realized with the H- (3He)-D three-ion heating ICRH scenario, foreseen for f ~ 25 MHz in W7-X. An ICRH system is prepared for W7-X, expected to deposit RF powers up to ~1.5 MW (depending on the coupling) at frequencies between 25-38 MHz in pulses up to 10s. A two-strap ICRH antenna for W7-X is under construction. Each strap of the antenna is on one side connected to a tuning capacitor and grounded to the antenna box at the other end. A prematching has been implemented by connecting the RF transmission lines at an intermediate position on each strap. The main dimensions of straps and antenna box have been optimized to maximise the power delivered to the plasma, using the reference plasma density profile in front of the antenna, provided by the W7-X team. A dedicated test stand is under construction to check the main functional tests of the full ICRH antenna system.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
An important test for W7-X is to demonstrate confinement of fast trapped ions at volume averaged beta values up to 5%, corresponding to plasma densities above 1020 m-3. Energetic ions in W7-X with energies 50 < E < 100 keV mimic alphas in a reactor. To generate such a population is a challenging task in high-density plasmas but this can be efficiently realized with the H- (3He)-D three-ion heating ICRH scenario, foreseen for f ~ 25 MHz in W7-X. An ICRH system is prepared for W7-X, expected to deposit RF powers up to ~1.5 MW (depending on the coupling) at frequencies between 25-38 MHz in pulses up to 10s. A two-strap ICRH antenna for W7-X is under construction. Each strap of the antenna is on one side connected to a tuning capacitor and grounded to the antenna box at the other end. A prematching has been implemented by connecting the RF transmission lines at an intermediate position on each strap. The main dimensions of straps and antenna box have been optimized to maximise the power delivered to the plasma, using the reference plasma density profile in front of the antenna, provided by the W7-X team. A dedicated test stand is under construction to check the main functional tests of the full ICRH antenna system.
R Otin; W Tierens; F Parra; S Aria; E. Lerche; P Jacquet; I Monakhov; P. Dumortier; B. Van Compernolle; JET Contributors
Full wave simulation of RF waves in cold plasma with the stabilized open-source finite element tool ERMES Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 050009, 2020.
@article{1980,
title = {Full wave simulation of RF waves in cold plasma with the stabilized open-source finite element tool ERMES},
author = {R Otin and W Tierens and F Parra and S Aria and E. Lerche and P Jacquet and I Monakhov and P. Dumortier and B. Van Compernolle and JET Contributors},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {050009},
abstract = {When RF waves are applied in tokamaks with metal walls, sheath rectification effects associated with the fields induced in the scrape-off layer (SOL) may lead to enhanced plasma-wall interactions (i.e. heat-loads in the limiters, RF-induced impurity sources) which can endanger the integrity of the machine and limit the RF power. Currently, some numerical tools are being used to simulate the RF antenna near fields in the presence of magnetized plasmas, but they have their limitations. Some neglect completely the interaction of the RF waves with the low-density plasma close to the antenna. Others take into account these interactions, but generate numerical spurious oscillations around the Lower Hybrid Resonance (LHR). Simplifications to reach convergence had also been tried (e.g. neglect gyrotropy, increase electron density to avoid the LHR), but the fields obtained with these simplifications can be very different to the real ones and this difference can affect the accuracy of derived magnitudes which use these fields as an input (e.g. sheath rectification effects). In this work we try to overcome all the limitations mentioned above by customizing the open-source finite element code ERMES. This code implements a finite element formulation which allows to simulate the near fields of the RF antenna in a continuous gyrotropic non-homogeneous media without limits in the minimum value of the plasma density and provides stable solutions even in the presence of the LHR. Benchmarking of this approach is underway and comparison against measurements, semi-analytical approaches and other codes will be presented.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
When RF waves are applied in tokamaks with metal walls, sheath rectification effects associated with the fields induced in the scrape-off layer (SOL) may lead to enhanced plasma-wall interactions (i.e. heat-loads in the limiters, RF-induced impurity sources) which can endanger the integrity of the machine and limit the RF power. Currently, some numerical tools are being used to simulate the RF antenna near fields in the presence of magnetized plasmas, but they have their limitations. Some neglect completely the interaction of the RF waves with the low-density plasma close to the antenna. Others take into account these interactions, but generate numerical spurious oscillations around the Lower Hybrid Resonance (LHR). Simplifications to reach convergence had also been tried (e.g. neglect gyrotropy, increase electron density to avoid the LHR), but the fields obtained with these simplifications can be very different to the real ones and this difference can affect the accuracy of derived magnitudes which use these fields as an input (e.g. sheath rectification effects). In this work we try to overcome all the limitations mentioned above by customizing the open-source finite element code ERMES. This code implements a finite element formulation which allows to simulate the near fields of the RF antenna in a continuous gyrotropic non-homogeneous media without limits in the minimum value of the plasma density and provides stable solutions even in the presence of the LHR. Benchmarking of this approach is underway and comparison against measurements, semi-analytical approaches and other codes will be presented.
R. Ragona; T Batal; J Hillairet; A. Messiaen; M Firdaouss; J -M Bernard; J. Ongena
Progress on the design of a DEMO high power ICRH travelling wave antenna mock-up to be tested on WEST Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 070014, 2020.
@article{1988,
title = {Progress on the design of a DEMO high power ICRH travelling wave antenna mock-up to be tested on WEST},
author = {R. Ragona and T Batal and J Hillairet and A. Messiaen and M Firdaouss and J -M Bernard and J. Ongena},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {070014},
abstract = {WEST represents the ideal device to test a reactor relevant TWA due to its long pulse capability and full metal wall configuration. Moreover, the already installed ICRF high power launchers will allow a direct comparison between a classical in-port antenna and the TWA. This paper reports the progress in the design of an actively cooled high power mock-up of the WEST TWA antenna that will be tested in the TITAN facility. The main objective of the test is to assess the voltage stand-off of the antenna at a power level relevant for future operation. In order to be installed and operated in WEST, the antenna design has to comply with the specific machine requirements, in particular active cooling, magnetic configuration and interface with existing auxiliary heating systems. Here several aspects are taken into account: toroidal magnetic field ripple, field line inter-connection and thermal loads on the antenna. An initial assessment of mechanical compatibility in case of VDE is performed. In conclusion, the next steps and implications of the project are outlined.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
WEST represents the ideal device to test a reactor relevant TWA due to its long pulse capability and full metal wall configuration. Moreover, the already installed ICRF high power launchers will allow a direct comparison between a classical in-port antenna and the TWA. This paper reports the progress in the design of an actively cooled high power mock-up of the WEST TWA antenna that will be tested in the TITAN facility. The main objective of the test is to assess the voltage stand-off of the antenna at a power level relevant for future operation. In order to be installed and operated in WEST, the antenna design has to comply with the specific machine requirements, in particular active cooling, magnetic configuration and interface with existing auxiliary heating systems. Here several aspects are taken into account: toroidal magnetic field ripple, field line inter-connection and thermal loads on the antenna. An initial assessment of mechanical compatibility in case of VDE is performed. In conclusion, the next steps and implications of the project are outlined.
I. Stepanov; V Borsuk; K. Crombé; Casta D n; G Offermanns; J. Ongena; B Schweer; M. Vervier
Measurements of the scattering matrix of the W7-X prototype ICRH antenna and capacitor units Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 040003, 2020.
@article{1975,
title = {Measurements of the scattering matrix of the W7-X prototype ICRH antenna and capacitor units},
author = {I. Stepanov and V Borsuk and K. Crombé and Casta D n and G Offermanns and J. Ongena and B Schweer and M. Vervier},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {040003},
abstract = {Low-power measurements of the W7-X prototype ICRH antenna and variable capacitor units have been performed at FZJ Jülich. The antenna scattering matrices, with and without the carbon limiter tiles, have been de-embedded from network analyzer measurements using custom-built RF adapters. The results have been compared to the matrices calculated with CST Microwave Studio 2017 using a model based on the original CAD files. A very good agreement is seen, since a) the difference in magnitude is less than 0.02 in the whole frequency range, and ~ 0.015 or less in the range of interest (between 20 and 40 MHz) and b) the difference in phase is less than 5◦ in the whole frequency range, and ~ 2◦ or less between 20 and 40 MHz.
The S-parameters of the tuning capacitors have also been measured, to determine the relationship between the internal DC capacitance CDC of each unit, and the actual ''equivalent'' capacitance Ceq seen by the antenna at the ports. It is seen that the latter, 36-475 pF, is substantially different from the nominal DC range, which is 15-200 pF. Nonetheless this is not expected to pose a problem for antenna operation.
The S-parameters of other key components of the ICRF system, such as the matching unit and line sections containing directional couplers, were also measured and, where applicable, compared to models or used to extract calibration data.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Low-power measurements of the W7-X prototype ICRH antenna and variable capacitor units have been performed at FZJ Jülich. The antenna scattering matrices, with and without the carbon limiter tiles, have been de-embedded from network analyzer measurements using custom-built RF adapters. The results have been compared to the matrices calculated with CST Microwave Studio 2017 using a model based on the original CAD files. A very good agreement is seen, since a) the difference in magnitude is less than 0.02 in the whole frequency range, and ~ 0.015 or less in the range of interest (between 20 and 40 MHz) and b) the difference in phase is less than 5◦ in the whole frequency range, and ~ 2◦ or less between 20 and 40 MHz.
The S-parameters of the tuning capacitors have also been measured, to determine the relationship between the internal DC capacitance CDC of each unit, and the actual ''equivalent'' capacitance Ceq seen by the antenna at the ports. It is seen that the latter, 36-475 pF, is substantially different from the nominal DC range, which is 15-200 pF. Nonetheless this is not expected to pose a problem for antenna operation.
The S-parameters of other key components of the ICRF system, such as the matching unit and line sections containing directional couplers, were also measured and, where applicable, compared to models or used to extract calibration data.
The S-parameters of the tuning capacitors have also been measured, to determine the relationship between the internal DC capacitance CDC of each unit, and the actual ''equivalent'' capacitance Ceq seen by the antenna at the ports. It is seen that the latter, 36-475 pF, is substantially different from the nominal DC range, which is 15-200 pF. Nonetheless this is not expected to pose a problem for antenna operation.
The S-parameters of other key components of the ICRF system, such as the matching unit and line sections containing directional couplers, were also measured and, where applicable, compared to models or used to extract calibration data.
G Urbanczyk; L Colas; X J Zhang; W Helou; Y P Zhao; J Hillairet; X Z Gong; E. Lerche; G Lombard; Q C Ming; M Goniche; Z Ling; P Mollard; V Bobkov; X D Yang; O Meyer; L N Lu; J Gunn; C Yan; C Desgranges; J M Bernard; T Zhang; F Clairet; Pégouri B é; D. Van Eester; R Dumont; W Tierens; F. Durodié; B Zhang; J G Li; EAST WEST Teams
ICRH coupling optimization and impurity behavior in EAST and WEST Journal Article
In: AIP Conference proceedings, vol. 2254, no. 1, pp. 030012, 2020.
@article{1974,
title = {ICRH coupling optimization and impurity behavior in EAST and WEST},
author = {G Urbanczyk and L Colas and X J Zhang and W Helou and Y P Zhao and J Hillairet and X Z Gong and E. Lerche and G Lombard and Q C Ming and M Goniche and Z Ling and P Mollard and V Bobkov and X D Yang and O Meyer and L N Lu and J Gunn and C Yan and C Desgranges and J M Bernard and T Zhang and F Clairet and Pégouri B é and D. Van Eester and R Dumont and W Tierens and F. Durodié and B Zhang and J G Li and EAST WEST Teams},
year = {2020},
date = {2020-09-01},
journal = {AIP Conference proceedings},
volume = {2254},
number = {1},
pages = {030012},
abstract = {This study compares experimental observations on the two challenges of Ion Cyclotron Resonance Heating (ICRH), namely the coupling of waves to the plasma and the enhanced impurity sputtering in EAST and WEST medium size tokamaks.
In WEST, Lower Hybrid (LH) power helps improving ICRH coupling. In both machines experiments reveal that fueling from the midplane not only helps to couple waves from nearby antennas like in other devices, but also has an impact on the scrape-off layer (SOL) density in regions that are not magnetically connected to the valves. Localized midplane nozzle valves allow similar (in WEST) or better (in EAST) coupling compared to poloidally distributed valves. Core density control requirements for long-pulse operation, in particular in L-mode regime, however limit the amount of gas that can be injected.
During ICRH impurities can contaminate the plasma up to a level detrimental for the operation, e.g. up to 100% of ICRH power can be radiated on WEST in certain high power conditions. In WEST, tungsten (W) production measured by visible spectroscopy increases on all the observed objects during ICRH compared to a reference phase without ICRH. On some components (antenna side limiters, baffle, divertor) the rise is larger than with a similar LH power. The relative contribution of each object and physical process (RF-sheaths, fast ion ripple losses) to core contamination yet remains poorly known. Comparing antenna limiters with W-coating vs low-Z materials would help quantifying the role of these components. In EAST, the core W content, measured by EUV spectroscopy in presence of divertor sources only, is correlated with the total injected power, either from ICRH or LH. Since 2018 the LH guard limiter tiles were W-coated. Their contribution to the core W content appears more important than divertor sources when two-strap ICRH antennas magnetically connected to W components at the midplane is powered, compromising high performance operations.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
This study compares experimental observations on the two challenges of Ion Cyclotron Resonance Heating (ICRH), namely the coupling of waves to the plasma and the enhanced impurity sputtering in EAST and WEST medium size tokamaks.
In WEST, Lower Hybrid (LH) power helps improving ICRH coupling. In both machines experiments reveal that fueling from the midplane not only helps to couple waves from nearby antennas like in other devices, but also has an impact on the scrape-off layer (SOL) density in regions that are not magnetically connected to the valves. Localized midplane nozzle valves allow similar (in WEST) or better (in EAST) coupling compared to poloidally distributed valves. Core density control requirements for long-pulse operation, in particular in L-mode regime, however limit the amount of gas that can be injected.
During ICRH impurities can contaminate the plasma up to a level detrimental for the operation, e.g. up to 100% of ICRH power can be radiated on WEST in certain high power conditions. In WEST, tungsten (W) production measured by visible spectroscopy increases on all the observed objects during ICRH compared to a reference phase without ICRH. On some components (antenna side limiters, baffle, divertor) the rise is larger than with a similar LH power. The relative contribution of each object and physical process (RF-sheaths, fast ion ripple losses) to core contamination yet remains poorly known. Comparing antenna limiters with W-coating vs low-Z materials would help quantifying the role of these components. In EAST, the core W content, measured by EUV spectroscopy in presence of divertor sources only, is correlated with the total injected power, either from ICRH or LH. Since 2018 the LH guard limiter tiles were W-coated. Their contribution to the core W content appears more important than divertor sources when two-strap ICRH antennas magnetically connected to W components at the midplane is powered, compromising high performance operations.
In WEST, Lower Hybrid (LH) power helps improving ICRH coupling. In both machines experiments reveal that fueling from the midplane not only helps to couple waves from nearby antennas like in other devices, but also has an impact on the scrape-off layer (SOL) density in regions that are not magnetically connected to the valves. Localized midplane nozzle valves allow similar (in WEST) or better (in EAST) coupling compared to poloidally distributed valves. Core density control requirements for long-pulse operation, in particular in L-mode regime, however limit the amount of gas that can be injected.
During ICRH impurities can contaminate the plasma up to a level detrimental for the operation, e.g. up to 100% of ICRH power can be radiated on WEST in certain high power conditions. In WEST, tungsten (W) production measured by visible spectroscopy increases on all the observed objects during ICRH compared to a reference phase without ICRH. On some components (antenna side limiters, baffle, divertor) the rise is larger than with a similar LH power. The relative contribution of each object and physical process (RF-sheaths, fast ion ripple losses) to core contamination yet remains poorly known. Comparing antenna limiters with W-coating vs low-Z materials would help quantifying the role of these components. In EAST, the core W content, measured by EUV spectroscopy in presence of divertor sources only, is correlated with the total injected power, either from ICRH or LH. Since 2018 the LH guard limiter tiles were W-coated. Their contribution to the core W content appears more important than divertor sources when two-strap ICRH antennas magnetically connected to W components at the midplane is powered, compromising high performance operations.
H Faugel; V Bobkov; H F ü; J. -M. Noterdaeme; A. Messiaen; D. Van Eester
ICRF system efficiency Journal Article
In: Fusion Engineering and Design, vol. 156, pp. 111641, 2020, ISSN: 0920-3796.
@article{1969,
title = {ICRF system efficiency},
author = {H Faugel and V Bobkov and H F ü and J. -M. Noterdaeme and A. Messiaen and D. Van Eester},
url = {http://www.sciencedirect.com/science/article/pii/S0920379620301897},
doi = {https://doi.org/10.1016/j.fusengdes.2020.111641},
issn = {0920-3796},
year = {2020},
date = {2020-07-01},
journal = {Fusion Engineering and Design},
volume = {156},
pages = {111641},
abstract = {The efficiency of heating and current drive systems is one of the key parameters for a successful operation of fusion demonstration power plants like DEMO. In an earlier review article, overall plug efficiencies of H & CD systems were estimated at 20 – 30 % Bradley et al.. We present here a detailed breakdown based where possible on experimental data for the overall efficiency (plug to power in plasma) of ICRF (ion cyclotron range of frequencies) systems: 1) the technical efficiencies (RF generator, transmission lines, losses in antenna); 2) the interface efficiency (hardware/ plasma) and 3) heating efficiency (absorption in plasma). This leads currently to an overall efficiency for heating in the range 40% to 55%. Future improvements can lead to an overall efficiency of up to 69 %. In a second step we address the current drive efficiency (in terms of kA/MW absorbed).},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The efficiency of heating and current drive systems is one of the key parameters for a successful operation of fusion demonstration power plants like DEMO. In an earlier review article, overall plug efficiencies of H & CD systems were estimated at 20 – 30 % Bradley et al.. We present here a detailed breakdown based where possible on experimental data for the overall efficiency (plug to power in plasma) of ICRF (ion cyclotron range of frequencies) systems: 1) the technical efficiencies (RF generator, transmission lines, losses in antenna); 2) the interface efficiency (hardware/ plasma) and 3) heating efficiency (absorption in plasma). This leads currently to an overall efficiency for heating in the range 40% to 55%. Future improvements can lead to an overall efficiency of up to 69 %. In a second step we address the current drive efficiency (in terms of kA/MW absorbed).
A. Messiaen; V. Maquet
Coaxial and surface mode excitation by an ICRF antenna in large machines like DEMO and ITER Journal Article
In: Nuclear Fusion, vol. 60, no. 7, pp. 076014, 2020.
@article{1970,
title = {Coaxial and surface mode excitation by an ICRF antenna in large machines like DEMO and ITER},
author = {A. Messiaen and V. Maquet},
url = {https://doi.org/10.1088%2F1741-4326%2Fab8d05},
doi = {10.1088/1741-4326/ab8d05},
year = {2020},
date = {2020-07-01},
journal = {Nuclear Fusion},
volume = {60},
number = {7},
pages = {076014},
publisher = {IOP Publishing},
abstract = {A study of radio frequency (RF) field excitation in the edge of the plasma of DEMO is performed by means of the semi-analytic coupling code ANTITER II. The modeling uses the designed antenna and a reference low coupling density profile of ITER. The results show the existence of coaxial modes propagating between the wall and the Alfvén resonance region where surface modes are excited leading to large standing wave patterns all around the machine. The excitation of these modes can be strongly reduced for a strap current distribution of the antenna array which fulfills the two conditions and (I
i
: current of strap i, S
i
: toroidal strap position). These conditions are satisfied by the triple strap antenna of AUG that has allowed successful ICRH operation in a W coated machine. They are also achieved for the (0π
π0) phasing of the four columns of two poloidal triplets of radiating straps constituting the ITER antenna array. Ways to further optimize the cancellation of these edge modes are also investigated.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
A study of radio frequency (RF) field excitation in the edge of the plasma of DEMO is performed by means of the semi-analytic coupling code ANTITER II. The modeling uses the designed antenna and a reference low coupling density profile of ITER. The results show the existence of coaxial modes propagating between the wall and the Alfvén resonance region where surface modes are excited leading to large standing wave patterns all around the machine. The excitation of these modes can be strongly reduced for a strap current distribution of the antenna array which fulfills the two conditions and (I
i
: current of strap i, S
i
: toroidal strap position). These conditions are satisfied by the triple strap antenna of AUG that has allowed successful ICRH operation in a W coated machine. They are also achieved for the (0π
π0) phasing of the four columns of two poloidal triplets of radiating straps constituting the ITER antenna array. Ways to further optimize the cancellation of these edge modes are also investigated.
i
: current of strap i, S
i
: toroidal strap position). These conditions are satisfied by the triple strap antenna of AUG that has allowed successful ICRH operation in a W coated machine. They are also achieved for the (0π
π0) phasing of the four columns of two poloidal triplets of radiating straps constituting the ITER antenna array. Ways to further optimize the cancellation of these edge modes are also investigated.
S A Bozhenkov; Y. O. Kazakov; O P Ford; M N A Beurskens; Alcus J ó; J A Alonso; J Baldzuhn; C Brandt; K J Brunner; H Damm; G Fuchert; J Geiger; O Grulke; M Hirsch; U H ö; Z Huang; J Knauer; M Krychowiak; A Langenberg; H P Laqua; S Lazerson; N B Marushchenko; D Moseev; M Otte; N Pablant; E Pasch; A Pavone; J H E Proll; K Rahbarnia; E R Scott; H M Smith; T Stange; A Stechow; H Thomsen; Yu. Turkin; G Wurden; P Xanthopoulos; D Zhang; Wolf R C and
High-performance plasmas after pellet injections in Wendelstein 7-X Journal Article
In: Nuclear Fusion, vol. 60, no. 6, pp. 066011, 2020.
@article{1964,
title = {High-performance plasmas after pellet injections in Wendelstein 7-X},
author = {S A Bozhenkov and Y. O. Kazakov and O P Ford and M N A Beurskens and Alcus J ó and J A Alonso and J Baldzuhn and C Brandt and K J Brunner and H Damm and G Fuchert and J Geiger and O Grulke and M Hirsch and U H ö and Z Huang and J Knauer and M Krychowiak and A Langenberg and H P Laqua and S Lazerson and N B Marushchenko and D Moseev and M Otte and N Pablant and E Pasch and A Pavone and J H E Proll and K Rahbarnia and E R Scott and H M Smith and T Stange and A Stechow and H Thomsen and Yu. Turkin and G Wurden and P Xanthopoulos and D Zhang and Wolf R C and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab7867},
doi = {10.1088/1741-4326/ab7867},
year = {2020},
date = {2020-06-01},
journal = {Nuclear Fusion},
volume = {60},
number = {6},
pages = {066011},
publisher = {IOP Publishing},
abstract = {A significant improvement of plasma parameters in the optimized stellarator W7-X is found after injections of frozen hydrogen pellets. The ion temperature in the post-pellet phase exceeds 3 keV with 5 MW of electron heating and the global energy confinement time surpasses the empirical ISS04-scaling. The plasma parameters realized in such experiments are significantly above those in comparable gas-fuelled discharges. In this paper, we present details of these pellet experiments and discuss the main plasma properties during the enhanced confinement phases. Local power balance is applied to show that the heat transport in post-pellet phases is close to the neoclassical level for the ion channel and is about a factor of two above that level for the combined losses. In comparable gas-fuelled discharges, the heat transport is by about ten times larger than the neoclassical level, and thus is largely anomalous. It is further observed that the improvement in the transport is related to the peaked density profiles that lead to a stabilization of the ion-scale turbulence.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
A significant improvement of plasma parameters in the optimized stellarator W7-X is found after injections of frozen hydrogen pellets. The ion temperature in the post-pellet phase exceeds 3 keV with 5 MW of electron heating and the global energy confinement time surpasses the empirical ISS04-scaling. The plasma parameters realized in such experiments are significantly above those in comparable gas-fuelled discharges. In this paper, we present details of these pellet experiments and discuss the main plasma properties during the enhanced confinement phases. Local power balance is applied to show that the heat transport in post-pellet phases is close to the neoclassical level for the ion channel and is about a factor of two above that level for the combined losses. In comparable gas-fuelled discharges, the heat transport is by about ten times larger than the neoclassical level, and thus is largely anomalous. It is further observed that the improvement in the transport is related to the peaked density profiles that lead to a stabilization of the ion-scale turbulence.
F J Casson; H Patten; C Bourdelle; S Breton; J Citrin; F Koechl; M Sertoli; C Angioni; Y Baranov; R Bilato; E A Belli; C D Challis; G Corrigan; A Czarnecka; O Ficker; L Frassinetti; L Garzotti; M Goniche; J P Graves; T Johnson; K Kirov; P Knight; E. Lerche; M Mantsinen; J Mylnar; Valisa M and
Predictive multi-channel flux-driven modelling to optimise ICRH tungsten control and fusion performance in JET Journal Article
In: Nuclear Fusion, vol. 60, no. 6, pp. 066029, 2020.
@article{1968,
title = {Predictive multi-channel flux-driven modelling to optimise ICRH tungsten control and fusion performance in JET},
author = {F J Casson and H Patten and C Bourdelle and S Breton and J Citrin and F Koechl and M Sertoli and C Angioni and Y Baranov and R Bilato and E A Belli and C D Challis and G Corrigan and A Czarnecka and O Ficker and L Frassinetti and L Garzotti and M Goniche and J P Graves and T Johnson and K Kirov and P Knight and E. Lerche and M Mantsinen and J Mylnar and Valisa M and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab833f},
doi = {10.1088/1741-4326/ab833f},
year = {2020},
date = {2020-06-01},
journal = {Nuclear Fusion},
volume = {60},
number = {6},
pages = {066029},
publisher = {IOP Publishing},
abstract = {The evolution of the JET high performance hybrid scenario, including central accumulation of the tungsten (W) impurity, is reproduced with predictive multi-channel integrated modelling over multiple confinement times using first-principle based core transport models. Eight transport channels () are modelled predictively, with self-consistent sources, radiation and magnetic equilibrium, yielding a system with multiple non-linearities: This system can reproduce the observed radiative temperature collapse after several confinement times. W is transported inward by neoclassical convection driven by the main ion density gradients and enhanced by poloidal asymmetries due to centrifugal acceleration. The slow evolution of the bulk density profile sets the timescale for W accumulation. Modelling this phenomenon requires a turbulent transport model capable of accurately predicting particle and momentum transport (QuaLiKiz) and a neoclassical transport model including the effects of poloidal asymmetries (NEO) coupled to an integrated plasma simulator (JINTRAC). The modelling capability is applied to optimise the available actuators to prevent W accumulation, and to extrapolate in power and pulse length. Central NBI heating is preferred for high performance, but gives central deposition of particles and torque which increase the risk of W accumulation by increasing density peaking and poloidal asymmetry. The primary mechanism for ICRH to control W in JET is via its impact through turbulence in reducing main ion density peaking (which drives inward neoclassical convection), increased temperature screening and turbulent W diffusion. The anisotropy from ICRH also reduces poloidal asymmetry, but this effect is negligible in high rotation JET discharges. High power ICRH near the axis can sensitively mitigate against W accumulation, and dominant ion heating (e.g. He-3 minority) is predicted to provide more resilience to W accumulation than dominant electron heating (e.g. H minority) in the JET hybrid scenario. Extrapolation to DT plasmas finds 17.5 MW of fusion power and improved confinement compared to DD, due to reduced ion-electron energy exchange, and increased Ti/Te stabilisation of ITG instabilities. The turbulence reduction in DT increases density peaking and accelerates the arrival of W on axis; this may be mitigated by reducing the penetration of the beam particle source with an increased pedestal density.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The evolution of the JET high performance hybrid scenario, including central accumulation of the tungsten (W) impurity, is reproduced with predictive multi-channel integrated modelling over multiple confinement times using first-principle based core transport models. Eight transport channels () are modelled predictively, with self-consistent sources, radiation and magnetic equilibrium, yielding a system with multiple non-linearities: This system can reproduce the observed radiative temperature collapse after several confinement times. W is transported inward by neoclassical convection driven by the main ion density gradients and enhanced by poloidal asymmetries due to centrifugal acceleration. The slow evolution of the bulk density profile sets the timescale for W accumulation. Modelling this phenomenon requires a turbulent transport model capable of accurately predicting particle and momentum transport (QuaLiKiz) and a neoclassical transport model including the effects of poloidal asymmetries (NEO) coupled to an integrated plasma simulator (JINTRAC). The modelling capability is applied to optimise the available actuators to prevent W accumulation, and to extrapolate in power and pulse length. Central NBI heating is preferred for high performance, but gives central deposition of particles and torque which increase the risk of W accumulation by increasing density peaking and poloidal asymmetry. The primary mechanism for ICRH to control W in JET is via its impact through turbulence in reducing main ion density peaking (which drives inward neoclassical convection), increased temperature screening and turbulent W diffusion. The anisotropy from ICRH also reduces poloidal asymmetry, but this effect is negligible in high rotation JET discharges. High power ICRH near the axis can sensitively mitigate against W accumulation, and dominant ion heating (e.g. He-3 minority) is predicted to provide more resilience to W accumulation than dominant electron heating (e.g. H minority) in the JET hybrid scenario. Extrapolation to DT plasmas finds 17.5 MW of fusion power and improved confinement compared to DD, due to reduced ion-electron energy exchange, and increased Ti/Te stabilisation of ITG instabilities. The turbulence reduction in DT increases density peaking and accelerates the arrival of W on axis; this may be mitigated by reducing the penetration of the beam particle source with an increased pedestal density.
D. Van Eester; E. Lerche
Solving the all-FLR ICRH integro-differential wave equation as a high-order di↵erential equation Technical Report
LPP-ERM/KMS no. 145, 2020.
@techreport{1966,
title = {Solving the all-FLR ICRH integro-differential wave equation as a high-order di↵erential equation},
author = {D. Van Eester and E. Lerche},
year = {2020},
date = {2020-06-01},
number = {145},
institution = {LPP-ERM/KMS},
keywords = {},
pubstate = {published},
tppubtype = {techreport}
}
S N Gerasimov; P Abreu; G Artaserse; M Baruzzo; P Buratti; I S Carvalho; I H Coffey; De La E Luna; T C Hender; R B Henriques; R Felton; S. Jachmich; U Kruezi; P J Lomas; P McCullen; M Maslov; E Matveeva; S. Moradi; L Piron; F G Rimini; W Schippers; C Stuart; G Szepesi; M Tsalas; D Valcarcel; Zakharov L E and
Overview of disruptions with JET-ILW Journal Article
In: Nuclear Fusion, vol. 60, no. 6, pp. 066028, 2020.
@article{1965,
title = {Overview of disruptions with JET-ILW},
author = {S N Gerasimov and P Abreu and G Artaserse and M Baruzzo and P Buratti and I S Carvalho and I H Coffey and De La E Luna and T C Hender and R B Henriques and R Felton and S. Jachmich and U Kruezi and P J Lomas and P McCullen and M Maslov and E Matveeva and S. Moradi and L Piron and F G Rimini and W Schippers and C Stuart and G Szepesi and M Tsalas and D Valcarcel and Zakharov L E and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab87b0},
doi = {10.1088/1741-4326/ab87b0},
year = {2020},
date = {2020-06-01},
journal = {Nuclear Fusion},
volume = {60},
number = {6},
pages = {066028},
publisher = {IOP Publishing},
abstract = {The paper presents an analysis of disruptions occurring during JET-ILW plasma operations covering the period from the start of ILW (ITER-like wall) operation up to completion of JET operation in 2016. The total number of disruptions was 1951 including 466 with deliberately induced disruptions. The average rate of unintended disruptions was 16.1 %, which is significantly above the ITER target at 15 MA. The pre-disruptive plasma parameters are: plasma current Ip = (0.82–3.38) MA, toroidal field BT = (0.98–3.4) T, safety factor q95 = (1.52–9.05), plasma internal inductance li = (0.58–1.86), Greenwald density limit fraction FGWL = (0.04–1.61), with 720 X-point plasma pulses from a subset of 1420 unintended disruption shots. Massive gas injection (MGI) has been routinely used in protection mode both to terminate pulses when the plasma is at risk of disruption and to mitigate against disruption effects. The MGI was mainly triggered by the n = 1 locked mode (LM) amplitude exceeding a threshold or by the disruption itself, namely, either dIp/dt (specifically, a fast drop in Ip
) or the toroidal loop voltage exceeding threshold values. For mitigation purposes, only the LM was used as a physics precursor and threshold on the LM signal was used to trigger the MGI prior to disruption. Long lasting LM (≥ 100 ms) do exist prior to disruption in 75% of cases. However, 10% of non-disruptive pulses have a LM which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D configurations, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated vertical displacement events (VDEs) generally have significant plasma current toroidal asymmetries. Unmitigated non-VDE disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during the CQ and so they too are subject to large vertical displacements. MGI is a reliable tool to mitigate 3D effects and correspondingly sideways forces during the CQ. The vessel structure loads depend on the force impulse and force time behaviour, including their rotation. The toroidal rotation of 3D configuration may cause resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The JET-ILW amplitude-frequency interdependence of toroidal rotation of 3D configurations is presented.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The paper presents an analysis of disruptions occurring during JET-ILW plasma operations covering the period from the start of ILW (ITER-like wall) operation up to completion of JET operation in 2016. The total number of disruptions was 1951 including 466 with deliberately induced disruptions. The average rate of unintended disruptions was 16.1 %, which is significantly above the ITER target at 15 MA. The pre-disruptive plasma parameters are: plasma current Ip = (0.82–3.38) MA, toroidal field BT = (0.98–3.4) T, safety factor q95 = (1.52–9.05), plasma internal inductance li = (0.58–1.86), Greenwald density limit fraction FGWL = (0.04–1.61), with 720 X-point plasma pulses from a subset of 1420 unintended disruption shots. Massive gas injection (MGI) has been routinely used in protection mode both to terminate pulses when the plasma is at risk of disruption and to mitigate against disruption effects. The MGI was mainly triggered by the n = 1 locked mode (LM) amplitude exceeding a threshold or by the disruption itself, namely, either dIp/dt (specifically, a fast drop in Ip
) or the toroidal loop voltage exceeding threshold values. For mitigation purposes, only the LM was used as a physics precursor and threshold on the LM signal was used to trigger the MGI prior to disruption. Long lasting LM (≥ 100 ms) do exist prior to disruption in 75% of cases. However, 10% of non-disruptive pulses have a LM which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D configurations, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated vertical displacement events (VDEs) generally have significant plasma current toroidal asymmetries. Unmitigated non-VDE disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during the CQ and so they too are subject to large vertical displacements. MGI is a reliable tool to mitigate 3D effects and correspondingly sideways forces during the CQ. The vessel structure loads depend on the force impulse and force time behaviour, including their rotation. The toroidal rotation of 3D configuration may cause resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The JET-ILW amplitude-frequency interdependence of toroidal rotation of 3D configurations is presented.
) or the toroidal loop voltage exceeding threshold values. For mitigation purposes, only the LM was used as a physics precursor and threshold on the LM signal was used to trigger the MGI prior to disruption. Long lasting LM (≥ 100 ms) do exist prior to disruption in 75% of cases. However, 10% of non-disruptive pulses have a LM which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D configurations, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated vertical displacement events (VDEs) generally have significant plasma current toroidal asymmetries. Unmitigated non-VDE disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during the CQ and so they too are subject to large vertical displacements. MGI is a reliable tool to mitigate 3D effects and correspondingly sideways forces during the CQ. The vessel structure loads depend on the force impulse and force time behaviour, including their rotation. The toroidal rotation of 3D configuration may cause resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The JET-ILW amplitude-frequency interdependence of toroidal rotation of 3D configurations is presented.
Y He; Yuhong Xu; I. Shesterikov; M. Vergote; A Kraemer-Flecken; J Cheng
Impact of edge magnetic perturbation (MP) on multi-scale turbulence and turbulent transport across a MP-induced edge transport reduction in the TEXTOR tokamak Journal Article
In: Physics of Plasmas, vol. 27, no. 6, pp. 062511, 2020.
@article{1967,
title = {Impact of edge magnetic perturbation (MP) on multi-scale turbulence and turbulent transport across a MP-induced edge transport reduction in the TEXTOR tokamak},
author = {Y He and Yuhong Xu and I. Shesterikov and M. Vergote and A Kraemer-Flecken and J Cheng},
year = {2020},
date = {2020-06-01},
journal = {Physics of Plasmas},
volume = {27},
number = {6},
pages = {062511},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
J Baldzuhn; H Damm; C D Beidler; K McCarthy; N Panadero; C Biedermann; S A Bozhenkov; A Dinklage; K J Brunner; G Fuchert; Y. O. Kazakov; M Beurskens; M Dibon; J Geiger; O Grulke; U H ö; T Klinger; F K ö; J Knauer; G Kocsis; P Kornejew; P T Lang; A Langenberg; H Laqua; N A Pablant; E Pasch; T S Pedersen; B Ploeckl; K Rahbarnia; G Schlisio; E R Scott; T Stange; Von A Stechow; T Szepesi; Y Turkin; F Wagner; V Winters; G Wurden; Zhang D and
Enhanced energy confinement after series of pellets in Wendelstein 7-X Journal Article
In: Plasma Physics and Controlled Fusion, vol. 62, no. 5, pp. 055012, 2020.
@article{1963,
title = {Enhanced energy confinement after series of pellets in Wendelstein 7-X},
author = {J Baldzuhn and H Damm and C D Beidler and K McCarthy and N Panadero and C Biedermann and S A Bozhenkov and A Dinklage and K J Brunner and G Fuchert and Y. O. Kazakov and M Beurskens and M Dibon and J Geiger and O Grulke and U H ö and T Klinger and F K ö and J Knauer and G Kocsis and P Kornejew and P T Lang and A Langenberg and H Laqua and N A Pablant and E Pasch and T S Pedersen and B Ploeckl and K Rahbarnia and G Schlisio and E R Scott and T Stange and Von A Stechow and T Szepesi and Y Turkin and F Wagner and V Winters and G Wurden and Zhang D and},
url = {https://doi.org/10.1088%2F1361-6587%2Fab8112},
doi = {10.1088/1361-6587/ab8112},
year = {2020},
date = {2020-05-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {62},
number = {5},
pages = {055012},
publisher = {IOP Publishing},
abstract = {A series of ice pellets was injected into the advanced stellarator Wendelstein 7-X (W7-X). Although the pellets were small and slow, deep and efficient particle fueling could be observed experimentally. The most striking feature appearing after the injection of the pellets, however, was a transient increase in the energy confinement time. This transient phase resembled in several aspects modes of enhanced confinement after gas-puff or pellet injection, as observed in other fusion experiments. All experimental attempts, to prolong this phase, failed. In this paper, discharges are described that show the enhanced energy confinement, and some conditions are summarized which seem to be essential in order to generate it. The focus here is on deep particle fueling by pellets, and shaping of the density profiles during and after the series of pellets. During this time, neutral gas particle re-fueling at the plasma edge is reduced, while density profile peaking and low impurity radiation losses are present.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
A series of ice pellets was injected into the advanced stellarator Wendelstein 7-X (W7-X). Although the pellets were small and slow, deep and efficient particle fueling could be observed experimentally. The most striking feature appearing after the injection of the pellets, however, was a transient increase in the energy confinement time. This transient phase resembled in several aspects modes of enhanced confinement after gas-puff or pellet injection, as observed in other fusion experiments. All experimental attempts, to prolong this phase, failed. In this paper, discharges are described that show the enhanced energy confinement, and some conditions are summarized which seem to be essential in order to generate it. The focus here is on deep particle fueling by pellets, and shaping of the density profiles during and after the series of pellets. During this time, neutral gas particle re-fueling at the plasma edge is reduced, while density profile peaking and low impurity radiation losses are present.
P Vallejos; T Johnson; R. Ragona; D. Van Eester; B Zaar; T Hellsten
Iterative addition of finite Larmor radius effects to finite element models using wavelet decomposition Journal Article
In: Plasma Physics and Controlled Fusion, vol. 62, no. 4, pp. 045022, 2020.
@article{1962,
title = {Iterative addition of finite Larmor radius effects to finite element models using wavelet decomposition},
author = {P Vallejos and T Johnson and R. Ragona and D. Van Eester and B Zaar and T Hellsten},
url = {https://doi.org/10.1088%2F1361-6587%2Fab6f55},
doi = {10.1088/1361-6587/ab6f55},
year = {2020},
date = {2020-04-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {62},
number = {4},
pages = {045022},
publisher = {IOP Publishing},
abstract = {Modeling the propagation and damping of electromagnetic waves in a hot magnetized plasma is difficult due to spatial dispersion. In such media, the dielectric response becomes non-local and the wave equation an integro-differential equation. In the application of RF heating and current drive in tokamak plasmas, the finite Larmor radius (FLR) causes spatial dispersion, which gives rise to physical phenomena such as higher harmonic ion cyclotron damping and mode conversion to electrostatic waves. In this paper, a new numerical method based on an iterative wavelet finite element scheme is presented, which is suitable for adding non-local effects to the wave equation by iterations. To verify the method, we apply it to a case of one-dimensional fast wave heating at the second harmonic ion cyclotron resonance, and study mode conversion to ion Bernstein waves (IBW) in a toroidal plasma. Comparison with a local (truncated FLR) model showed good agreement in general. The observed difference is in the damping of the IBW, where the proposed method predicts stronger damping on the IBW.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Modeling the propagation and damping of electromagnetic waves in a hot magnetized plasma is difficult due to spatial dispersion. In such media, the dielectric response becomes non-local and the wave equation an integro-differential equation. In the application of RF heating and current drive in tokamak plasmas, the finite Larmor radius (FLR) causes spatial dispersion, which gives rise to physical phenomena such as higher harmonic ion cyclotron damping and mode conversion to electrostatic waves. In this paper, a new numerical method based on an iterative wavelet finite element scheme is presented, which is suitable for adding non-local effects to the wave equation by iterations. To verify the method, we apply it to a case of one-dimensional fast wave heating at the second harmonic ion cyclotron resonance, and study mode conversion to ion Bernstein waves (IBW) in a toroidal plasma. Comparison with a local (truncated FLR) model showed good agreement in general. The observed difference is in the damping of the IBW, where the proposed method predicts stronger damping on the IBW.
G Fuchert; K J Brunner; K Rahbarnia; T Stange; D Zhang; J Baldzuhn; S A Bozhenkov; C D Beidler; M N A Beurskens; S Brezinsek; R Burhenn; H Damm; A Dinklage; Y Feng; P Hacker; M Hirsch; Y. O. Kazakov; J Knauer; A Langenberg; H P Laqua; S Lazerson; N A Pablant; E Pasch; F Reimold; Sunn T Pedersen; E R Scott; F Warmer; V R Winters; Wolf R C and
Increasing the density in Wendelstein 7-X: benefits and limitations Journal Article
In: Nuclear Fusion, vol. 60, no. 3, pp. 036020, 2020.
@article{1958,
title = {Increasing the density in Wendelstein 7-X: benefits and limitations},
author = {G Fuchert and K J Brunner and K Rahbarnia and T Stange and D Zhang and J Baldzuhn and S A Bozhenkov and C D Beidler and M N A Beurskens and S Brezinsek and R Burhenn and H Damm and A Dinklage and Y Feng and P Hacker and M Hirsch and Y. O. Kazakov and J Knauer and A Langenberg and H P Laqua and S Lazerson and N A Pablant and E Pasch and F Reimold and Sunn T Pedersen and E R Scott and F Warmer and V R Winters and Wolf R C and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab6d40},
doi = {10.1088/1741-4326/ab6d40},
year = {2020},
date = {2020-03-01},
journal = {Nuclear Fusion},
volume = {60},
number = {3},
pages = {036020},
publisher = {IOP Publishing},
abstract = {In stellarators, increasing the density is beneficial for the energy confinement. While there is no single reason for this observation, it is still very robust across different devices and this is reflected in the empirical energy confinement time scaling for stellarators, ISS04. In order to study whether this is also true for Wendelstein 7-X, the density scaling of the energy confinement time is analyzed and compared to ISS04 for the first divertor experiments. When the density is increased beyond a critical density, however, radiative collapses are frequently observed. Existing analytical models for the critical density are revisited to assess whether they can predict the accessible density range. Furthermore, since close to the collapse the radiation losses increase substantially, the impact on the global energy confinement is investigated. It is found that in plasmas with high radiation the density scaling of the energy confinement time becomes weaker, the reason for this observation is not yet clear. In the second half of the first divertor campaign, boronization was applied to W7-X for the first time. This broadened the operational window, allowing for operation at higher density and, hence, higher stored energy.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
In stellarators, increasing the density is beneficial for the energy confinement. While there is no single reason for this observation, it is still very robust across different devices and this is reflected in the empirical energy confinement time scaling for stellarators, ISS04. In order to study whether this is also true for Wendelstein 7-X, the density scaling of the energy confinement time is analyzed and compared to ISS04 for the first divertor experiments. When the density is increased beyond a critical density, however, radiative collapses are frequently observed. Existing analytical models for the critical density are revisited to assess whether they can predict the accessible density range. Furthermore, since close to the collapse the radiation losses increase substantially, the impact on the global energy confinement is investigated. It is found that in plasmas with high radiation the density scaling of the energy confinement time becomes weaker, the reason for this observation is not yet clear. In the second half of the first divertor campaign, boronization was applied to W7-X for the first time. This broadened the operational window, allowing for operation at higher density and, hence, higher stored energy.
A. Goriaev; T. Wauters; R Brakel; S Brezinsek; A Dinklage; J Fellinger; H Grote; D Moseev; S Sereda; Volzke O and
Wall conditioning at the Wendelstein 7-X stellarator operating with a graphite divertor Journal Article
In: Physica Scripta, vol. 2020, no. T171, pp. 014063, 2020.
@article{1961,
title = {Wall conditioning at the Wendelstein 7-X stellarator operating with a graphite divertor},
author = {A. Goriaev and T. Wauters and R Brakel and S Brezinsek and A Dinklage and J Fellinger and H Grote and D Moseev and S Sereda and Volzke O and},
url = {https://doi.org/10.1088%2F1402-4896%2Fab60fa},
doi = {10.1088/1402-4896/ab60fa},
year = {2020},
date = {2020-03-01},
journal = {Physica Scripta},
volume = {2020},
number = {T171},
pages = {014063},
publisher = {IOP Publishing},
abstract = {Wall conditioning plays an important role in achieving record plasma performance on the superconducting stellarator Wendelstein 7-X (W7-X) by controlling plasma density and maintaining low levels of impurities. The development of the optimal wall conditioning strategy is of high importance for successful operation of the W7-X with a graphite divertor. The current strategy consists of initial wall conditioning applied prior to a physics experimental program and wall conditioning techniques executed during the plasma operation phase. The combination of baking and glow discharge cleaning provides a fast start to reliable plasma operation. Regular boronisation opens a new operational window of high plasma densities above 1 * 1020 m−3. Strong wall fueling remains, however, one of the main issues preventing plasma density control. A reliable solution is provided by application of He electron cyclotron resonance heating pulse trains, which are optimised to maximise fuel removal.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Wall conditioning plays an important role in achieving record plasma performance on the superconducting stellarator Wendelstein 7-X (W7-X) by controlling plasma density and maintaining low levels of impurities. The development of the optimal wall conditioning strategy is of high importance for successful operation of the W7-X with a graphite divertor. The current strategy consists of initial wall conditioning applied prior to a physics experimental program and wall conditioning techniques executed during the plasma operation phase. The combination of baking and glow discharge cleaning provides a fast start to reliable plasma operation. Regular boronisation opens a new operational window of high plasma densities above 1 * 1020 m−3. Strong wall fueling remains, however, one of the main issues preventing plasma density control. A reliable solution is provided by application of He electron cyclotron resonance heating pulse trains, which are optimised to maximise fuel removal.
S. Moradi; A Thyagaraja
Impact of Micro-Scale Stochastic Zonal Flows on the Macro-Scale Visco-Resistive Magnetohydrodynamic Modes Journal Article
In: Mathematics, vol. 8, no. 3, pp. 443, 2020.
@article{1956,
title = {Impact of Micro-Scale Stochastic Zonal Flows on the Macro-Scale Visco-Resistive Magnetohydrodynamic Modes},
author = {S. Moradi and A Thyagaraja},
year = {2020},
date = {2020-03-01},
journal = {Mathematics},
volume = {8},
number = {3},
pages = {443},
abstract = {A model is developed to simulate micro-scale turbulence driven Zonal Flows (ZFs), and their impact on the Magnetohydrodynamic (MHD) tearing and kink modes is examined. The model is based on a stochastic representation of the micro-scale ZFs with a given Alfvén Mach number, MS. Two approaches were explored: (i) passive stochastic model where the ZFs amplitudes are independent of the MHD mode amplitude, and (ii) the semi-stochastic model where the amplitudes of the ZFs have a dependence on the amplitude of the MHD mode itself. The results show that the stochastic ZFs can significantly stabilise the (2,1) and (1,1) MHD modes even at very low kinematic viscosity, where the mode is linearly unstable. Our results therefore indicate a possible mechanism for stabilisation of the MHD modes via small-scale perturbations in poloidal flow, simulating the turbulence driven ZFs.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
A model is developed to simulate micro-scale turbulence driven Zonal Flows (ZFs), and their impact on the Magnetohydrodynamic (MHD) tearing and kink modes is examined. The model is based on a stochastic representation of the micro-scale ZFs with a given Alfvén Mach number, MS. Two approaches were explored: (i) passive stochastic model where the ZFs amplitudes are independent of the MHD mode amplitude, and (ii) the semi-stochastic model where the amplitudes of the ZFs have a dependence on the amplitude of the MHD mode itself. The results show that the stochastic ZFs can significantly stabilise the (2,1) and (1,1) MHD modes even at very low kinematic viscosity, where the mode is linearly unstable. Our results therefore indicate a possible mechanism for stabilisation of the MHD modes via small-scale perturbations in poloidal flow, simulating the turbulence driven ZFs.
N Pablant; A Langenberg; A Alonso; J Baldzuhn; C D Beidler; S Bozhenkov; R Burhenn; K J Brunner; A Dinklage; G Fuchert; O Ford; D A Gates; J Geiger; M Hirsch; U H ö; Y. O. Kazakov; J Knauer; M Krychowiak; H Laqua; M Landreman; S Lazerson; Maa H ß; O Marchuck; A Mollen; E Pasch; A Pavone; S Satake; Schr T ö; H M Smith; J Svensson; P Traverso; Y Turkin; J L Velasco; A Stechow; F Warmer; G Weir; R C Wolf; Zhang D and
Investigation of the neoclassical ambipolar electric field in ion-root plasmas on W7-X Journal Article
In: Nuclear Fusion, vol. 60, no. 3, pp. 036021, 2020.
@article{1959,
title = {Investigation of the neoclassical ambipolar electric field in ion-root plasmas on W7-X},
author = {N Pablant and A Langenberg and A Alonso and J Baldzuhn and C D Beidler and S Bozhenkov and R Burhenn and K J Brunner and A Dinklage and G Fuchert and O Ford and D A Gates and J Geiger and M Hirsch and U H ö and Y. O. Kazakov and J Knauer and M Krychowiak and H Laqua and M Landreman and S Lazerson and Maa H ß and O Marchuck and A Mollen and E Pasch and A Pavone and S Satake and Schr T ö and H M Smith and J Svensson and P Traverso and Y Turkin and J L Velasco and A Stechow and F Warmer and G Weir and R C Wolf and Zhang D and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab6ea8},
doi = {10.1088/1741-4326/ab6ea8},
year = {2020},
date = {2020-03-01},
journal = {Nuclear Fusion},
volume = {60},
number = {3},
pages = {036021},
publisher = {IOP Publishing},
abstract = {The role of the radial electric field in high-performance ion-root plasmas on Wendelstein 7-X (W7-X) is examined and compared with neoclassical predictions. The W7-X stellarator is the world's first large-scale optimized stellarator. One of the important targets chosen for optimization during the W7-X design process was the reduction of core neoclassical heat transport. This optimization was targeted for reactor-relevant high-density plasmas with in which the neoclassical ambipolar radial electric field is expected to be negative throughout the plasmas core.
Measurements of the core radial electric field (E
r
) have confirmed that ion-root conditions (negative E
r
in the plasma core) have been achieved in W7-X with high-density plasmas and central ECRH. These measured E
r
profiles agree well with the neoclassical ambipolar E
r
predicted by the code sfincs. This good agreement provides confidence in the validity of neoclassical calculations in high-density ion-root conditions, and enables initial studies on the role of neoclassical transport in the optimized high-density regime of W7-X.
Profile measurements of electron temperature (T
e
), ion temperature (T
i
) and electron density (n
e
) along with approximations for the average value of have been used as inputs to the sfincs code to calculate the ambipolar E
r
profile along with neoclassical ion and electron energy flux profiles (, ). Finally the total experimental energy input to the plasma from ECRH heating has been compared to the neoclassical energy fluxes to provide a first estimate for the fraction of transport that can be attributed to neoclassical processes in reactor-relevant high-density ion-root conditions.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The role of the radial electric field in high-performance ion-root plasmas on Wendelstein 7-X (W7-X) is examined and compared with neoclassical predictions. The W7-X stellarator is the world's first large-scale optimized stellarator. One of the important targets chosen for optimization during the W7-X design process was the reduction of core neoclassical heat transport. This optimization was targeted for reactor-relevant high-density plasmas with in which the neoclassical ambipolar radial electric field is expected to be negative throughout the plasmas core.
Measurements of the core radial electric field (E
r
) have confirmed that ion-root conditions (negative E
r
in the plasma core) have been achieved in W7-X with high-density plasmas and central ECRH. These measured E
r
profiles agree well with the neoclassical ambipolar E
r
predicted by the code sfincs. This good agreement provides confidence in the validity of neoclassical calculations in high-density ion-root conditions, and enables initial studies on the role of neoclassical transport in the optimized high-density regime of W7-X.
Profile measurements of electron temperature (T
e
), ion temperature (T
i
) and electron density (n
e
) along with approximations for the average value of have been used as inputs to the sfincs code to calculate the ambipolar E
r
profile along with neoclassical ion and electron energy flux profiles (, ). Finally the total experimental energy input to the plasma from ECRH heating has been compared to the neoclassical energy fluxes to provide a first estimate for the fraction of transport that can be attributed to neoclassical processes in reactor-relevant high-density ion-root conditions.
Measurements of the core radial electric field (E
r
) have confirmed that ion-root conditions (negative E
r
in the plasma core) have been achieved in W7-X with high-density plasmas and central ECRH. These measured E
r
profiles agree well with the neoclassical ambipolar E
r
predicted by the code sfincs. This good agreement provides confidence in the validity of neoclassical calculations in high-density ion-root conditions, and enables initial studies on the role of neoclassical transport in the optimized high-density regime of W7-X.
Profile measurements of electron temperature (T
e
), ion temperature (T
i
) and electron density (n
e
) along with approximations for the average value of have been used as inputs to the sfincs code to calculate the ambipolar E
r
profile along with neoclassical ion and electron energy flux profiles (, ). Finally the total experimental energy input to the plasma from ECRH heating has been compared to the neoclassical energy fluxes to provide a first estimate for the fraction of transport that can be attributed to neoclassical processes in reactor-relevant high-density ion-root conditions.
E Wang; S Brezinsek; S Sereda; Buttensch B ö; T Barbui; C P Dhard; M Endler; O Ford; E Flom; K C Hammond; M Jakubowski; M Krychowiak; P Kornejew; R K ö; Y Liang; M Mayer; D Naujoks; O Neubauer; J Oelmann; M Rasinski; V R Winters; A. Goriaev; T. Wauters; Y Wei; Zhang D and
Impurity sources and fluxes in W7-X: from the plasma-facing components to the edge layer Journal Article
In: Physica Scripta, vol. 2020, no. T171, pp. 014040, 2020.
@article{1960,
title = {Impurity sources and fluxes in W7-X: from the plasma-facing components to the edge layer},
author = {E Wang and S Brezinsek and S Sereda and Buttensch B ö and T Barbui and C P Dhard and M Endler and O Ford and E Flom and K C Hammond and M Jakubowski and M Krychowiak and P Kornejew and R K ö and Y Liang and M Mayer and D Naujoks and O Neubauer and J Oelmann and M Rasinski and V R Winters and A. Goriaev and T. Wauters and Y Wei and Zhang D and},
url = {https://doi.org/10.1088%2F1402-4896%2Fab4c04},
doi = {10.1088/1402-4896/ab4c04},
year = {2020},
date = {2020-03-01},
journal = {Physica Scripta},
volume = {2020},
number = {T171},
pages = {014040},
publisher = {IOP Publishing},
abstract = {Wendelstein 7-X (W7-X) is a nearly full-carbon machine with graphite divertors, baffles and shields in Operation Phase 1.2b (OP 1.2b). Divertor spectrometer measurements showed that an amount of helium and oxygen impurities existed in the predominately hydrogen plasma, which resulted in a high carbon impurity level by enhanced physical and chemical sputtering by these impurities in comparison with the pure impinging proton yields. In order to improve the wall conditions, especially to reduce the oxygen content, boronizations were applied in OP1.2b. After the boronization, an oxygen decrease by more than an order of magnitude was observed. Helium disappeared in comparison with OP1.2a due to reduced application of helium wall conditioning after introduction of boronizations. The overall radiation normalized to line integrated density was reduced by a factor of six. In addition, local CH4 injection was applied in the divertor in order to quantify the chemical sputtering by hydrogen on divertor plates. The experimentally determined effective D/XB of the A–X band of CH resulting from CH4 was at T
e ≈ 20 eV and n
e ≈ 5 × 1018 m−3. It was applied to determine the hydrocarbon fluxes and further to deduce the particle flux ratio ГCH4/ГH on divertor plates.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Wendelstein 7-X (W7-X) is a nearly full-carbon machine with graphite divertors, baffles and shields in Operation Phase 1.2b (OP 1.2b). Divertor spectrometer measurements showed that an amount of helium and oxygen impurities existed in the predominately hydrogen plasma, which resulted in a high carbon impurity level by enhanced physical and chemical sputtering by these impurities in comparison with the pure impinging proton yields. In order to improve the wall conditions, especially to reduce the oxygen content, boronizations were applied in OP1.2b. After the boronization, an oxygen decrease by more than an order of magnitude was observed. Helium disappeared in comparison with OP1.2a due to reduced application of helium wall conditioning after introduction of boronizations. The overall radiation normalized to line integrated density was reduced by a factor of six. In addition, local CH4 injection was applied in the divertor in order to quantify the chemical sputtering by hydrogen on divertor plates. The experimentally determined effective D/XB of the A–X band of CH resulting from CH4 was at T
e ≈ 20 eV and n
e ≈ 5 × 1018 m−3. It was applied to determine the hydrocarbon fluxes and further to deduce the particle flux ratio ГCH4/ГH on divertor plates.
e ≈ 20 eV and n
e ≈ 5 × 1018 m−3. It was applied to determine the hydrocarbon fluxes and further to deduce the particle flux ratio ГCH4/ГH on divertor plates.
T. Wauters; D Borodin; R Brakel; S Brezinsek; K J Brunner; J Buermans; S Coda; A Dinklage; D Douai; O Ford; G Fuchert; A. Goriaev; H Grote; A Hakola; E Joffrin; J Knauer; T Loarer; H Laqua; A. I. Lyssoivan; V Moiseenko; D Moseev; J. Ongena; K Rahbarnia; D Ricci; V Rohde; S Romanelli; S Sereda; T Stange; Tabar F L é; Lilla Van ó; O Volzke; E Wang; and; and
Wall conditioning in fusion devices with superconducting coils Journal Article
In: Plasma Physics and Controlled Fusion, vol. 62, no. 3, pp. 034002, 2020.
@article{1957,
title = {Wall conditioning in fusion devices with superconducting coils},
author = {T. Wauters and D Borodin and R Brakel and S Brezinsek and K J Brunner and J Buermans and S Coda and A Dinklage and D Douai and O Ford and G Fuchert and A. Goriaev and H Grote and A Hakola and E Joffrin and J Knauer and T Loarer and H Laqua and A. I. Lyssoivan and V Moiseenko and D Moseev and J. Ongena and K Rahbarnia and D Ricci and V Rohde and S Romanelli and S Sereda and T Stange and Tabar F L é and Lilla Van ó and O Volzke and E Wang and and and and},
url = {https://doi.org/10.1088%2F1361-6587%2Fab5ad0},
doi = {10.1088/1361-6587/ab5ad0},
year = {2020},
date = {2020-03-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {62},
number = {3},
pages = {034002},
publisher = {IOP Publishing},
abstract = {Wall conditioning is essential in tokamak and stellarator research to achieve plasma performance and reproducibility. This paper presents an overview of recent conditioning results, both from experiments in present devices and modelling, in view of devices with superconducting coils, with focus on W7-X, JT-60SA and ITER. In these devices, the coils stay energised throughout an experimental day or week which demands for new conditioning techniques that work in presence of the nominal field, in addition to the proven conditioning methods such as baking, glow discharge conditioning (GDC) and low-Z wall coating through GDC-plasma, which do not work under such condition. The discussed techniques are RF conditioning without plasma current, both in the ion cyclotron and electron cyclotron range of frequencies, and diverted conditioning plasmas with nested magnetic flux surfaces. Similarities and differences between tokamaks and stellarators are highlighted. Finally a conditional tritium recovery strategy for ITER is proposed based on Ion Cyclotron Wall Conditioning and L-mode plasma results from JET, equipped with an ITER-like wall (beryllium main chamber wall and tungsten divertor).},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Wall conditioning is essential in tokamak and stellarator research to achieve plasma performance and reproducibility. This paper presents an overview of recent conditioning results, both from experiments in present devices and modelling, in view of devices with superconducting coils, with focus on W7-X, JT-60SA and ITER. In these devices, the coils stay energised throughout an experimental day or week which demands for new conditioning techniques that work in presence of the nominal field, in addition to the proven conditioning methods such as baking, glow discharge conditioning (GDC) and low-Z wall coating through GDC-plasma, which do not work under such condition. The discussed techniques are RF conditioning without plasma current, both in the ion cyclotron and electron cyclotron range of frequencies, and diverted conditioning plasmas with nested magnetic flux surfaces. Similarities and differences between tokamaks and stellarators are highlighted. Finally a conditional tritium recovery strategy for ITER is proposed based on Ion Cyclotron Wall Conditioning and L-mode plasma results from JET, equipped with an ITER-like wall (beryllium main chamber wall and tungsten divertor).
P. Dumortier; E. Lerche; F. Durodié; T Blackman; W Helou; I Monakhov; C Noble
Review of the JET ILA scattering-matrix arc detection system Journal Article
In: Fusion Engineering and Design, vol. 150, pp. 110669, 2020, ISSN: 0920-3796.
@article{1954,
title = {Review of the JET ILA scattering-matrix arc detection system},
author = {P. Dumortier and E. Lerche and F. Durodié and T Blackman and W Helou and I Monakhov and C Noble},
url = {http://www.sciencedirect.com/science/article/pii/S0920379619300948},
doi = {https://doi.org/10.1016/j.fusengdes.2019.01.085},
issn = {0920-3796},
year = {2020},
date = {2020-01-01},
journal = {Fusion Engineering and Design},
volume = {150},
pages = {110669},
abstract = {Arc detection is an essential protection system for high power RF systems. It is commonly realised by monitoring the Voltage Standing Wave Ratio (VSWR) in the transmission lines. The JET ILA is a load-tolerant ICRF antenna with a low impedance section, for which the standard VSWR protection is ineffective. The Scattering-Matrix Arc Detection (SMAD) was proposed and implemented on JET to protect the low-impedance section around the T-junction against arcing. It is based on a consistency check of the RF signals around this section using a table of coefficients obtained from RF modelling. The paper reviews the SMAD principle, its commissioning on JET and the latest improvements to the system. It shows an example of operation and the time response of the system is analysed. The latest improvements have increased the robustness and reliability of SMAD, making it a choice of arc detection system for critical sections on existing and future RF systems.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Arc detection is an essential protection system for high power RF systems. It is commonly realised by monitoring the Voltage Standing Wave Ratio (VSWR) in the transmission lines. The JET ILA is a load-tolerant ICRF antenna with a low impedance section, for which the standard VSWR protection is ineffective. The Scattering-Matrix Arc Detection (SMAD) was proposed and implemented on JET to protect the low-impedance section around the T-junction against arcing. It is based on a consistency check of the RF signals around this section using a table of coefficients obtained from RF modelling. The paper reviews the SMAD principle, its commissioning on JET and the latest improvements to the system. It shows an example of operation and the time response of the system is analysed. The latest improvements have increased the robustness and reliability of SMAD, making it a choice of arc detection system for critical sections on existing and future RF systems.
G Giruzzi; M Yoshida; N Aiba; J F Artaud; J Ayllon-Guerola; O Beeke; A Bierwage; T Bolzonella; M Bonotto; C Boulbe; M Chernyshova; S Coda; R Coelho; D Corona; N Cruz; S Davis; C Day; De G Tommasi; M Dibon; D Douai; D Farina; A Fassina; B Faugeras; L Figini; M Fukumoto; S Futatani; K Galazka; J Garcia; Garcia-Mu M ñ; L Garzotti; L Giudicotti; N Hayashi; M Honda; K Hoshino; A Iantchenko; S Ide; S Inoue; A Isayama; E Joffrin; Y Kamada; K Kamiya; M Kashiwagi; H Kawashima; T Kobayashi; A Kojima; T Kurki-Suonio; P Lang; Ph Lauber; E Luna; G Marchiori; G Matsunaga; A Matsuyama; M Mattei; S Mazzi; A Mele; Y Miyata; S Moriyama; J Morales; A Moro; T Nakano; R Neu; S Nowak; FP Orsitto; V Ostuni; N Oyama; Pam S é; R Pasqualotto; Pégouri B é; E Perelli; L Pigatto; C Piron; A Pironti; P Platania; B Ploeckl; D Ricci; M Romanelli; G Rubino; S Sakurai; Särkim K ä; M Scannapiego; K Shinohara; J Shiraishi; S Soare; C Sozzi; T Suzuki; Y Suzuki; T Szepesi; M Takechi; K Tanaka; H Tojo; M Turnyanskiy; H Urano; M Valisa; M Vallar; J Varje; J Vega; F Villone; T Wakatsuki; T. Wauters; M Wischmeier; S Yamoto; Zag R ó
Advances in the physics studies for the JT-60SA tokamak exploitation and research plan Journal Article
In: Plasma Physics and Controlled Fusion, vol. 62, no. 1, pp. 014009, 2020.
@article{1950,
title = {Advances in the physics studies for the JT-60SA tokamak exploitation and research plan},
author = {G Giruzzi and M Yoshida and N Aiba and J F Artaud and J Ayllon-Guerola and O Beeke and A Bierwage and T Bolzonella and M Bonotto and C Boulbe and M Chernyshova and S Coda and R Coelho and D Corona and N Cruz and S Davis and C Day and De G Tommasi and M Dibon and D Douai and D Farina and A Fassina and B Faugeras and L Figini and M Fukumoto and S Futatani and K Galazka and J Garcia and Garcia-Mu M ñ and L Garzotti and L Giudicotti and N Hayashi and M Honda and K Hoshino and A Iantchenko and S Ide and S Inoue and A Isayama and E Joffrin and Y Kamada and K Kamiya and M Kashiwagi and H Kawashima and T Kobayashi and A Kojima and T Kurki-Suonio and P Lang and Ph Lauber and E Luna and G Marchiori and G Matsunaga and A Matsuyama and M Mattei and S Mazzi and A Mele and Y Miyata and S Moriyama and J Morales and A Moro and T Nakano and R Neu and S Nowak and FP Orsitto and V Ostuni and N Oyama and Pam S é and R Pasqualotto and Pégouri B é and E Perelli and L Pigatto and C Piron and A Pironti and P Platania and B Ploeckl and D Ricci and M Romanelli and G Rubino and S Sakurai and Särkim K ä and M Scannapiego and K Shinohara and J Shiraishi and S Soare and C Sozzi and T Suzuki and Y Suzuki and T Szepesi and M Takechi and K Tanaka and H Tojo and M Turnyanskiy and H Urano and M Valisa and M Vallar and J Varje and J Vega and F Villone and T Wakatsuki and T. Wauters and M Wischmeier and S Yamoto and Zag R ó},
url = {https://doi.org/10.1088%2F1361-6587%2Fab4771},
doi = {10.1088/1361-6587/ab4771},
year = {2020},
date = {2020-01-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {62},
number = {1},
pages = {014009},
publisher = {IOP Publishing},
abstract = {JT-60SA, the largest tokamak that will operate before ITER, has been designed and built jointly by Japan and Europe, and is due to start operation in 2020. Its main missions are to support ITER exploitation and to contribute to the demonstration fusion reactor machine and scenario design. Peculiar properties of JT-60SA are its capability to produce long-pulse, high-β, and highly shaped plasmas. The preparation of the JT-60SA Research Plan, plasma scenarios, and exploitation are producing physics results that are not only relevant to future JT-60SA experiments, but often constitute original contributions to plasma physics and fusion research. Results of this kind are presented in this paper, in particular in the areas of fast ion physics, high-beta plasma properties and control, and non-linear edge localised mode stability studies.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
JT-60SA, the largest tokamak that will operate before ITER, has been designed and built jointly by Japan and Europe, and is due to start operation in 2020. Its main missions are to support ITER exploitation and to contribute to the demonstration fusion reactor machine and scenario design. Peculiar properties of JT-60SA are its capability to produce long-pulse, high-β, and highly shaped plasmas. The preparation of the JT-60SA Research Plan, plasma scenarios, and exploitation are producing physics results that are not only relevant to future JT-60SA experiments, but often constitute original contributions to plasma physics and fusion research. Results of this kind are presented in this paper, in particular in the areas of fast ion physics, high-beta plasma properties and control, and non-linear edge localised mode stability studies.
C C Klepper; B Pegourie; S Vartanian; M Goniche; E Delabie; D. Van Eester; E. Lerche; G Sips; I Borodkina; D Douai; I Jepu; U Kruezi; G F Matthews; A Widdowson; and
Sub-divertor fuel isotopic content detection limit for JET and its impact on ICRF core heating and DTE2 operation Journal Article
In: Nuclear Fusion, vol. 60, no. 1, pp. 016021, 2020.
@article{1951,
title = {Sub-divertor fuel isotopic content detection limit for JET and its impact on ICRF core heating and DTE2 operation},
author = {C C Klepper and B Pegourie and S Vartanian and M Goniche and E Delabie and D. Van Eester and E. Lerche and G Sips and I Borodkina and D Douai and I Jepu and U Kruezi and G F Matthews and A Widdowson and and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab4c5a},
doi = {10.1088/1741-4326/ab4c5a},
year = {2020},
date = {2020-01-01},
journal = {Nuclear Fusion},
volume = {60},
number = {1},
pages = {016021},
publisher = {IOP Publishing},
abstract = {The ability to detect and control fuel isotopic content down to a 1% concentration level is greatly important for the upcoming JET DTE2 campaign, as well as its associated TT and DD phases. A reduction of H minority concentration even from 2% down to 1% is shown here to have significant impact on the effectiveness of ion cyclotron range of frequencies core heating, while the ability to maintain T or D concentration at or below 1% is critical to limiting fusion neutron generation in the DD and TT phases, correspondingly. The sub-divertor measurement of (global) isotopic concentration, based on Penning-activated optical spectroscopy, can deliver minimally this 1% detection for DTE2 as long as light collection from the Penning emission can be optimized and gradual window transmission deterioration can be minimized. This is simulated with a statistical analysis developed to understand the uncertainty sources in the JET DTE1 data, as well as to guide the optimization of an upgraded, fuel-isotopic content (and helium-ash concentration) gas analysis system for the JET divertor in preparation for DTE2. While this random error can be reduced to allow measurement substantially below 1% concentration, analysis also shows a systematic error of up to 1% understood to be due to plasma–surface interactions in the Penning excitation, suggesting that 1% may still be the low-end limit for the sub-divertor measurement, unless a Penning-source conditioning approach is also developed.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The ability to detect and control fuel isotopic content down to a 1% concentration level is greatly important for the upcoming JET DTE2 campaign, as well as its associated TT and DD phases. A reduction of H minority concentration even from 2% down to 1% is shown here to have significant impact on the effectiveness of ion cyclotron range of frequencies core heating, while the ability to maintain T or D concentration at or below 1% is critical to limiting fusion neutron generation in the DD and TT phases, correspondingly. The sub-divertor measurement of (global) isotopic concentration, based on Penning-activated optical spectroscopy, can deliver minimally this 1% detection for DTE2 as long as light collection from the Penning emission can be optimized and gradual window transmission deterioration can be minimized. This is simulated with a statistical analysis developed to understand the uncertainty sources in the JET DTE1 data, as well as to guide the optimization of an upgraded, fuel-isotopic content (and helium-ash concentration) gas analysis system for the JET divertor in preparation for DTE2. While this random error can be reduced to allow measurement substantially below 1% concentration, analysis also shows a systematic error of up to 1% understood to be due to plasma–surface interactions in the Penning excitation, suggesting that 1% may still be the low-end limit for the sub-divertor measurement, unless a Penning-source conditioning approach is also developed.
M Nocente; Dal A Molin; N Eidietis; L Giacomelli; G Gorini; Y. O. Kazakov; E Khilkevitch; V Kiptily; M Iliasova; A Lvovskiy; M Mantsinen; A Mariani; E Panontin; G Papp; G Pautasso; C Paz-Soldan; D Rigamonti; M Salewski; A Shevelev; M Tardocchi; and
MeV range particle physics studies in tokamak plasmas using gamma-ray spectroscopy Journal Article
In: Plasma Physics and Controlled Fusion, vol. 62, no. 1, pp. 014015, 2020.
@article{1952,
title = {MeV range particle physics studies in tokamak plasmas using gamma-ray spectroscopy},
author = {M Nocente and Dal A Molin and N Eidietis and L Giacomelli and G Gorini and Y. O. Kazakov and E Khilkevitch and V Kiptily and M Iliasova and A Lvovskiy and M Mantsinen and A Mariani and E Panontin and G Papp and G Pautasso and C Paz-Soldan and D Rigamonti and M Salewski and A Shevelev and M Tardocchi and and},
url = {https://doi.org/10.1088%2F1361-6587%2Fab4f32},
doi = {10.1088/1361-6587/ab4f32},
year = {2020},
date = {2020-01-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {62},
number = {1},
pages = {014015},
publisher = {IOP Publishing},
abstract = {Gamma-ray spectroscopy (GRS) has become an established technique to determine properties of the distribution function of the energetic particles in the MeV range, which are fast ions from heating and fusion reactions or runaway electrons born in disruptions. In this paper we present a selection of recent results where GRS is key to investigate the physics of MeV range particles. These range from radio-frequency heating experiments, where theoretical models can be tested with an unprecedented degree of accuracy, to disruption mitigation studies, where GRS sheds light on the effect of the actuators on the runaway electron velocity space. We further discuss the unique observational capabilities offered by the technique in deuterium–tritium plasmas, particularly with regard to the inference of the energy- and pitch-resolved distribution function of the α particles born from fusion reactions in the plasma core.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Gamma-ray spectroscopy (GRS) has become an established technique to determine properties of the distribution function of the energetic particles in the MeV range, which are fast ions from heating and fusion reactions or runaway electrons born in disruptions. In this paper we present a selection of recent results where GRS is key to investigate the physics of MeV range particles. These range from radio-frequency heating experiments, where theoretical models can be tested with an unprecedented degree of accuracy, to disruption mitigation studies, where GRS sheds light on the effect of the actuators on the runaway electron velocity space. We further discuss the unique observational capabilities offered by the technique in deuterium–tritium plasmas, particularly with regard to the inference of the energy- and pitch-resolved distribution function of the α particles born from fusion reactions in the plasma core.
R. Ragona; A. Messiaen; J. Ongena; D. Van Eester; M. Van Schoor; J -M Bernard; J Hillairet; J. -M. Noterdaeme
A travelling wave array system as solution for the ion cyclotron resonance frequencies heating of DEMO Journal Article
In: Nuclear Fusion, vol. 60, no. 1, pp. 016027, 2020.
@article{1953,
title = {A travelling wave array system as solution for the ion cyclotron resonance frequencies heating of DEMO},
author = {R. Ragona and A. Messiaen and J. Ongena and D. Van Eester and M. Van Schoor and J -M Bernard and J Hillairet and J. -M. Noterdaeme},
url = {https://doi.org/10.1088%2F1741-4326%2Fab504a},
doi = {10.1088/1741-4326/ab504a},
year = {2020},
date = {2020-01-01},
journal = {Nuclear Fusion},
volume = {60},
number = {1},
pages = {016027},
publisher = {IOP Publishing},
abstract = {Travelling wave array (TWA) antennas distributed along the periphery of the tokamak are presently considered as an ion cyclotron resonance frequencies (ICRF) heating solution for the DEMO reactor. Compared to the conventional ICRF antenna systems currently in use or designed for future machines like ITER, the TWA consists of antenna sections integrated in the breeding blanket scattered around the machine, each one fed through a variable coupler in a resonant ring configuration. Previous modelling of an antenna system for DEMO with 16 quadruple TWA sections of eight straps shows that a power capability exceeding 50 MW can be obtained in the frequency band of interest using the reference low coupling plasma profile of ITER. The described system optimizes the coupling to the plasma by providing a large number of radiating elements, which results in enhanced antenna directivity, hereby decreasing the antenna power density. This results in a maximum strap voltage amplitude of only 15 kV and maximum inter-strap voltage amplitude of 18 kV. The generators remain matched for all loading conditions: the system is totally load resilient. Following the recommendation of the work package heating and current drive Review Panel, a TWA ion cyclotron resonant heating (ICRH) system consisting of fewer sections concentrated in front of the equatorial ports is analysed in this paper and compared to the previous design. Reducing the number of sections increases the power density and its associated voltages. To couple 50 MW on the ITER density profile, voltages up to 30 kV are now required. Some aspects like the coupling between sections and its repercussion on the feeding network are briefly discussed. To assess the feasibility of the TWA fed by a resonant ring as an ICRH system for a DEMO reactor, a test on an existing medium size tokamak is under study.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Travelling wave array (TWA) antennas distributed along the periphery of the tokamak are presently considered as an ion cyclotron resonance frequencies (ICRF) heating solution for the DEMO reactor. Compared to the conventional ICRF antenna systems currently in use or designed for future machines like ITER, the TWA consists of antenna sections integrated in the breeding blanket scattered around the machine, each one fed through a variable coupler in a resonant ring configuration. Previous modelling of an antenna system for DEMO with 16 quadruple TWA sections of eight straps shows that a power capability exceeding 50 MW can be obtained in the frequency band of interest using the reference low coupling plasma profile of ITER. The described system optimizes the coupling to the plasma by providing a large number of radiating elements, which results in enhanced antenna directivity, hereby decreasing the antenna power density. This results in a maximum strap voltage amplitude of only 15 kV and maximum inter-strap voltage amplitude of 18 kV. The generators remain matched for all loading conditions: the system is totally load resilient. Following the recommendation of the work package heating and current drive Review Panel, a TWA ion cyclotron resonant heating (ICRH) system consisting of fewer sections concentrated in front of the equatorial ports is analysed in this paper and compared to the previous design. Reducing the number of sections increases the power density and its associated voltages. To couple 50 MW on the ITER density profile, voltages up to 30 kV are now required. Some aspects like the coupling between sections and its repercussion on the feeding network are briefly discussed. To assess the feasibility of the TWA fed by a resonant ring as an ICRH system for a DEMO reactor, a test on an existing medium size tokamak is under study.
2019
A Kostic; K. Crombé; R Dux; M Griener; R Ochoukov; I. Shesterikov; C Suarez; M Usoltceva; R. Casagrande; E H Martin; J. -M. Noterdaeme
Polarization Stark spectroscopy for spatially resolved measurements of electric fields in the sheaths of ICRF antenna Journal Article
In: Review of Scientific Instruments, vol. 90, no. 12, pp. 123101, 2019.
@article{1949,
title = {Polarization Stark spectroscopy for spatially resolved measurements of electric fields in the sheaths of ICRF antenna},
author = {A Kostic and K. Crombé and R Dux and M Griener and R Ochoukov and I. Shesterikov and C Suarez and M Usoltceva and R. Casagrande and E H Martin and J. -M. Noterdaeme},
year = {2019},
date = {2019-12-01},
journal = {Review of Scientific Instruments},
volume = {90},
number = {12},
pages = {123101},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
M Schneider; A R Polevoi; S H Kim; A Loarte; S D Pinches; J-F Artaud; E Militello-ASP; B Beaumont; R Bilato; D Boilson; D J Campbell; P. Dumortier; D Farina; L Figini; Y Gribov; M Henderson; R R Khayrutdinov; A A Kavin; F K ö; T Kurki-Suonio; A Kuyanov; P. U. Lamalle; E. Lerche; V E Lukash; A. Messiaen; V Parail; Särkim K ä; A Snicker; D. Van Eester
Modelling one-third field operation in the ITER pre-fusion power operation phase Journal Article
In: Nuclear Fusion, vol. 59, no. 12, pp. 126014, 2019.
@article{1948,
title = {Modelling one-third field operation in the ITER pre-fusion power operation phase},
author = {M Schneider and A R Polevoi and S H Kim and A Loarte and S D Pinches and J-F Artaud and E Militello-ASP and B Beaumont and R Bilato and D Boilson and D J Campbell and P. Dumortier and D Farina and L Figini and Y Gribov and M Henderson and R R Khayrutdinov and A A Kavin and F K ö and T Kurki-Suonio and A Kuyanov and P. U. Lamalle and E. Lerche and V E Lukash and A. Messiaen and V Parail and Särkim K ä and A Snicker and D. Van Eester},
url = {https://doi.org/10.1088%2F1741-4326%2Fab3de0},
doi = {10.1088/1741-4326/ab3de0},
year = {2019},
date = {2019-12-01},
journal = {Nuclear Fusion},
volume = {59},
number = {12},
pages = {126014},
publisher = {IOP Publishing},
abstract = {In the four-stage approach of the new ITER Research Plan, the first pre-fusion power operation (PFPO) phase will only have limited power available from external heating and current drive (H&CD) systems: 20–30 MW provided by the electron cyclotron resonance heating (ECRH) system. Accessing the H-mode confinement regime at such low auxiliary power requires operating at low magnetic field, plasma current and density, i.e. 1.8 T and 5 MA for a density between 40% and 50% of the Greenwald density. H-mode plasmas at 5 MA/1.8 T will also be investigated in the second PFPO phase when ITER will have its full complement of H&CD capabilities installed, i.e. 20–30 MW of ECRH, 20 MW of ion cyclotron resonance heating and 33 MW of neutral beam injection. This paper describes the operational constraints and the H&CD capabilities for such scenarios in hydrogen and helium plasmas, to assess their viability and the issues it will be possible to address with them. The modelling results show that 5 MA/1.8 T scenarios are viable and will allow the exploration of the H-mode physics and control issues foreseen in the ITER Research Programme in the PFPO phases.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
In the four-stage approach of the new ITER Research Plan, the first pre-fusion power operation (PFPO) phase will only have limited power available from external heating and current drive (H&CD) systems: 20–30 MW provided by the electron cyclotron resonance heating (ECRH) system. Accessing the H-mode confinement regime at such low auxiliary power requires operating at low magnetic field, plasma current and density, i.e. 1.8 T and 5 MA for a density between 40% and 50% of the Greenwald density. H-mode plasmas at 5 MA/1.8 T will also be investigated in the second PFPO phase when ITER will have its full complement of H&CD capabilities installed, i.e. 20–30 MW of ECRH, 20 MW of ion cyclotron resonance heating and 33 MW of neutral beam injection. This paper describes the operational constraints and the H&CD capabilities for such scenarios in hydrogen and helium plasmas, to assess their viability and the issues it will be possible to address with them. The modelling results show that 5 MA/1.8 T scenarios are viable and will allow the exploration of the H-mode physics and control issues foreseen in the ITER Research Programme in the PFPO phases.
E J Strait; J L Barr; M Baruzzo; J W Berkery; R J Buttery; P C Vries; N W Eidietis; R S Granetz; J M Hanson; C T Holcomb; D A Humphreys; J H Kim; E Kolemen; M Kong; M J Lanctot; M Lehnen; E. Lerche; N C Logan; M Maraschek; M Okabayashi; J K Park; A Pau; G Pautasso; F M Poli; C Rea; S A Sabbagh; O Sauter; E Schuster; U A Sheikh; C Sozzi; F Turco; A D Turnbull; Z R Wang; W P Wehner; L Zeng
Progress in disruption prevention for ITER Journal Article
In: Nuclear Fusion, vol. 59, no. 11, pp. 112012, 2019.
@article{1942,
title = {Progress in disruption prevention for ITER},
author = {E J Strait and J L Barr and M Baruzzo and J W Berkery and R J Buttery and P C Vries and N W Eidietis and R S Granetz and J M Hanson and C T Holcomb and D A Humphreys and J H Kim and E Kolemen and M Kong and M J Lanctot and M Lehnen and E. Lerche and N C Logan and M Maraschek and M Okabayashi and J K Park and A Pau and G Pautasso and F M Poli and C Rea and S A Sabbagh and O Sauter and E Schuster and U A Sheikh and C Sozzi and F Turco and A D Turnbull and Z R Wang and W P Wehner and L Zeng},
url = {https://doi.org/10.1088%2F1741-4326%2Fab15de},
doi = {10.1088/1741-4326/ab15de},
year = {2019},
date = {2019-11-01},
journal = {Nuclear Fusion},
volume = {59},
number = {11},
pages = {112012},
publisher = {IOP Publishing},
abstract = {Key plasma physics and real-time control elements needed for robustly stable operation of high fusion power discharges in ITER have been demonstrated in recent research worldwide. Recent analysis has identified the current density profile as the main drive for disruptive instabilities in discharges simulating ITER's baseline scenario with high and low external torque. Ongoing development of model-based profile control and active control of magnetohydrodynamic instabilities is improving the stability of multiple scenarios. Significant advances have been made toward real-time physics-based prediction of instabilities, including path-oriented analysis, active sensing, and machine learning techniques for prediction that are beginning to go beyond simple disruption mitigation trigger applications. Active intervention contributes to prevention of disruptions, including forced rotation of magnetic islands to prevent wall locking, and localized heating/current drive to shrink the islands. Stable discharge rampdowns have been achieved with the fastest ITER-like scaled current ramp rates, while maintaining an X-point configuration. These elements are being integrated into stable operating scenarios and new event-handling systems for off-normal events in order to develop the physics basis and techniques for robust control in ITER.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Key plasma physics and real-time control elements needed for robustly stable operation of high fusion power discharges in ITER have been demonstrated in recent research worldwide. Recent analysis has identified the current density profile as the main drive for disruptive instabilities in discharges simulating ITER's baseline scenario with high and low external torque. Ongoing development of model-based profile control and active control of magnetohydrodynamic instabilities is improving the stability of multiple scenarios. Significant advances have been made toward real-time physics-based prediction of instabilities, including path-oriented analysis, active sensing, and machine learning techniques for prediction that are beginning to go beyond simple disruption mitigation trigger applications. Active intervention contributes to prevention of disruptions, including forced rotation of magnetic islands to prevent wall locking, and localized heating/current drive to shrink the islands. Stable discharge rampdowns have been achieved with the fastest ITER-like scaled current ramp rates, while maintaining an X-point configuration. These elements are being integrated into stable operating scenarios and new event-handling systems for off-normal events in order to develop the physics basis and techniques for robust control in ITER.
M Usoltceva; R Ochoukov; W Tierens; A Kostic; K. Crombé; S Heuraux; J. -M. Noterdaeme
Simulation of the ion cyclotron range of frequencies slow wave and the lower hybrid resonance in 3D in RAPLICASOL Journal Article
In: Plasma Physics and Controlled Fusion, vol. 61, no. 11, pp. 115011, 2019.
@article{1947,
title = {Simulation of the ion cyclotron range of frequencies slow wave and the lower hybrid resonance in 3D in RAPLICASOL},
author = {M Usoltceva and R Ochoukov and W Tierens and A Kostic and K. Crombé and S Heuraux and J. -M. Noterdaeme},
url = {https://doi.org/10.1088%2F1361-6587%2Fab476d},
doi = {10.1088/1361-6587/ab476d},
year = {2019},
date = {2019-11-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {61},
number = {11},
pages = {115011},
publisher = {IOP Publishing},
abstract = {Ion cyclotron range of frequencies (ICRF) wave propagation is calculated theoretically for tokamak conditions and for linear magnetized plasma device IShTAR which is dedicated to the RF sheath studies. Only the slow wave (SW) mode of ICRF waves can propagate and be studied in IShTAR. Therefore it is possible to decouple the role of the different ICRF modes in the RF sheath effects. Numerical simulations of the ICRF SW are done in COMSOL in the framework of the existing cold plasma modelling package RAPLICASOL and the SW is for the first time modelled in 3D. To date, RAPLICASOL existed as a 3D wave coupling modelling approach which targets the fast wave (FW). Plasma is implemented as a material with manually assigned physical properties and a perfectly matched layer (PML) is used to absorb the wave energy. Here it is demonstarted how to adjust the RAPLICASOL PML for models with propagating SW. Field structures in the resonance cone shape obtained for the SW differ significantly from the FW and exhibit strong dependence on the density profile in the close proximity of the antenna. The lower-hybrid (LH) resonance is a constant issue in the attempts to model the SW. In this work an approach to obtain correct numerical solutions in the LH resonance presence is demonstrated. Results of this work can be used to improve the complex tokamak ICRF simulations, where so far the SW propagation on the edge has been avoided.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Ion cyclotron range of frequencies (ICRF) wave propagation is calculated theoretically for tokamak conditions and for linear magnetized plasma device IShTAR which is dedicated to the RF sheath studies. Only the slow wave (SW) mode of ICRF waves can propagate and be studied in IShTAR. Therefore it is possible to decouple the role of the different ICRF modes in the RF sheath effects. Numerical simulations of the ICRF SW are done in COMSOL in the framework of the existing cold plasma modelling package RAPLICASOL and the SW is for the first time modelled in 3D. To date, RAPLICASOL existed as a 3D wave coupling modelling approach which targets the fast wave (FW). Plasma is implemented as a material with manually assigned physical properties and a perfectly matched layer (PML) is used to absorb the wave energy. Here it is demonstarted how to adjust the RAPLICASOL PML for models with propagating SW. Field structures in the resonance cone shape obtained for the SW differ significantly from the FW and exhibit strong dependence on the density profile in the close proximity of the antenna. The lower-hybrid (LH) resonance is a constant issue in the attempts to model the SW. In this work an approach to obtain correct numerical solutions in the LH resonance presence is demonstrated. Results of this work can be used to improve the complex tokamak ICRF simulations, where so far the SW propagation on the edge has been avoided.
D. Van Eester; E. Lerche; R. Ragona; A. Messiaen; T. Wauters
Ion cyclotron resonance heating scenarios for DEMO Journal Article
In: Nuclear Fusion, vol. 59, no. 10, pp. 106051, 2019.
@article{1941,
title = {Ion cyclotron resonance heating scenarios for DEMO},
author = {D. Van Eester and E. Lerche and R. Ragona and A. Messiaen and T. Wauters},
url = {https://doi.org/10.1088%2F1741-4326%2Fab318b},
doi = {10.1088/1741-4326/ab318b},
year = {2019},
date = {2019-10-01},
journal = {Nuclear Fusion},
volume = {59},
number = {10},
pages = {106051},
publisher = {IOP Publishing},
abstract = {The present paper offers an overview of the potential of ion cyclotron resonance heating (ICRH) or radio frequency heating for the DEMO machine. It is found that various suitable heating schemes are available. Similar to ITER and in view of the limited bandwidth of about 10 MHz that can be achieved to ensure optimal functioning of the launcher, it is proposed to make core second harmonic tritium heating the key ion heating scheme, assisted by fundamental cyclotron heating 3He in the early phase of the discharge; for the present design of DEMO—with a static magnetic field strength of B o = 5.855 T—that places the T and 3He layers in the core for f = 60 MHz and suggests centering the bandwidth around that main operating frequency. In line with earlier studies for hot, dense plasmas in large-size magnetic confinement machines, it is shown that good single pass absorption is achieved but that the size as well as the operating density and temperature of the machine cause the electrons to absorb a non-negligible fraction of the power away from the core when core ion heating is aimed at. Current drive and alternative heating options are briefly discussed and a dedicated computation is done for the traveling wave antenna, proposed for DEMO in view of its compatibility with substantial antenna–plasma distances. The various tasks that ICRH can fulfill are briefly listed. Finally, the impact of transport and the sensitivity of the obtained results to changes in the machine parameters is commented on.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The present paper offers an overview of the potential of ion cyclotron resonance heating (ICRH) or radio frequency heating for the DEMO machine. It is found that various suitable heating schemes are available. Similar to ITER and in view of the limited bandwidth of about 10 MHz that can be achieved to ensure optimal functioning of the launcher, it is proposed to make core second harmonic tritium heating the key ion heating scheme, assisted by fundamental cyclotron heating 3He in the early phase of the discharge; for the present design of DEMO—with a static magnetic field strength of B o = 5.855 T—that places the T and 3He layers in the core for f = 60 MHz and suggests centering the bandwidth around that main operating frequency. In line with earlier studies for hot, dense plasmas in large-size magnetic confinement machines, it is shown that good single pass absorption is achieved but that the size as well as the operating density and temperature of the machine cause the electrons to absorb a non-negligible fraction of the power away from the core when core ion heating is aimed at. Current drive and alternative heating options are briefly discussed and a dedicated computation is done for the traveling wave antenna, proposed for DEMO in view of its compatibility with substantial antenna–plasma distances. The various tasks that ICRH can fulfill are briefly listed. Finally, the impact of transport and the sensitivity of the obtained results to changes in the machine parameters is commented on.
M Valovic; Y Baranov; A Boboc; J Buchanan; J Citrin; E Delabie; L Frassinetti; J M Fontdecaba; L Garzotti; C Giroud; R McKean; E. Lerche; V Kiptily; F K ö; M Marin; M Maslov; S Menmuir; G Tvalashvili; Weisen H and
Control of the hydrogen:deuterium isotope mixture using pellets in JET Journal Article
In: Nuclear Fusion, vol. 59, no. 10, pp. 106047, 2019.
@article{1940,
title = {Control of the hydrogen:deuterium isotope mixture using pellets in JET},
author = {M Valovic and Y Baranov and A Boboc and J Buchanan and J Citrin and E Delabie and L Frassinetti and J M Fontdecaba and L Garzotti and C Giroud and R McKean and E. Lerche and V Kiptily and F K ö and M Marin and M Maslov and S Menmuir and G Tvalashvili and Weisen H and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab3812},
doi = {10.1088/1741-4326/ab3812},
year = {2019},
date = {2019-10-01},
journal = {Nuclear Fusion},
volume = {59},
number = {10},
pages = {106047},
publisher = {IOP Publishing},
abstract = {Deuterium pellets are injected into an initially pure hydrogen H-mode plasma in order to control the hydrogen:deuterium (H:D) isotope mixture. The pellets are deposited in the outer 20% of the minor radius, similar to that expected in ITER, creating transiently hollow electron density profiles. A H:D isotope mixture of approximately 45%:55% is obtained in the core with a pellet fuelling throughput of similar to previous pellet fuelling experiments in pure deuterium. Evolution of the H:D mix in the core is reproduced using a simple model, although deuterium transport could be higher at the beginning of the pellet train compared with the flat-top phase.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Deuterium pellets are injected into an initially pure hydrogen H-mode plasma in order to control the hydrogen:deuterium (H:D) isotope mixture. The pellets are deposited in the outer 20% of the minor radius, similar to that expected in ITER, creating transiently hollow electron density profiles. A H:D isotope mixture of approximately 45%:55% is obtained in the core with a pellet fuelling throughput of similar to previous pellet fuelling experiments in pure deuterium. Evolution of the H:D mix in the core is reproduced using a simple model, although deuterium transport could be higher at the beginning of the pellet train compared with the flat-top phase.
Julien Hillairet; R. Ragona; Laurent Colas; Walid Helou; Fréd é
Lower hybrid range cold magnetized plasma coupling in ANSYS HFSS Journal Article
In: Fusion Engineering and Design, vol. 146, pp. 1473 - 1475, 2019, ISSN: 0920-3796, (SI:SOFT-30).
@article{1938,
title = {Lower hybrid range cold magnetized plasma coupling in ANSYS HFSS},
author = {Julien Hillairet and R. Ragona and Laurent Colas and Walid Helou and Fréd é},
url = {http://www.sciencedirect.com/science/article/pii/S092037961930290X},
doi = {https://doi.org/10.1016/j.fusengdes.2019.02.108},
issn = {0920-3796},
year = {2019},
date = {2019-09-01},
journal = {Fusion Engineering and Design},
volume = {146},
pages = {1473 - 1475},
abstract = {The coupling between cold magnetized plasmas with lower hybrid resonance frequency antennas is generally addressed using specifically developed codes. These antenna coupling codes often approximate the plasma to a surface impedance described by a 1D half infinite models and antennas either with simplified 2D dimensions or from 3D CAD models conversion. Such approaches add an additional step of model approximation/conversion which is not convenient to rapidly assess impact of geometry changes on coupling performances. In this work, we assess the ability of using ANSYS HFSS to describe the usual range of experimental magnetized cold plasma inhomogeneous parameters facing LHRF antennas, away from resonances, in order to guide the RF designer during the design phase of an antenna. The coupling calculations of Lower Hybrid Resonance Frequency antennas are performed and successfully benchmarked with the fast coupling codes ALOHA on inhomogeneous cold magnetized plasmas. Good agreements are obtained when the boundary conditions in the full-wave modelling are properly handled. Practical advices and limitations are given in order to define correctly these boundary conditions and insure correct results.},
note = {SI:SOFT-30},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The coupling between cold magnetized plasmas with lower hybrid resonance frequency antennas is generally addressed using specifically developed codes. These antenna coupling codes often approximate the plasma to a surface impedance described by a 1D half infinite models and antennas either with simplified 2D dimensions or from 3D CAD models conversion. Such approaches add an additional step of model approximation/conversion which is not convenient to rapidly assess impact of geometry changes on coupling performances. In this work, we assess the ability of using ANSYS HFSS to describe the usual range of experimental magnetized cold plasma inhomogeneous parameters facing LHRF antennas, away from resonances, in order to guide the RF designer during the design phase of an antenna. The coupling calculations of Lower Hybrid Resonance Frequency antennas are performed and successfully benchmarked with the fast coupling codes ALOHA on inhomogeneous cold magnetized plasmas. Good agreements are obtained when the boundary conditions in the full-wave modelling are properly handled. Practical advices and limitations are given in order to define correctly these boundary conditions and insure correct results.
J. -M. Noterdaeme; A. Messiaen; R. Ragona; W Zhang; A Bader; F. Durodié; U Fischer; T Franke; E Smigelskis; J. Ongena; M Q Tran; D. Van Eester; M. Van Schoor
Progress on an ion cyclotron range of frequency system for DEMO Journal Article
In: Fusion Engineering and Design, vol. 146, pp. 1321-1324, 2019, ISSN: 0920-3796, (SI:SOFT-30).
@article{1937,
title = {Progress on an ion cyclotron range of frequency system for DEMO},
author = {J. -M. Noterdaeme and A. Messiaen and R. Ragona and W Zhang and A Bader and F. Durodié and U Fischer and T Franke and E Smigelskis and J. Ongena and M Q Tran and D. Van Eester and M. Van Schoor},
url = {http://www.sciencedirect.com/science/article/pii/S0920379619302443},
doi = {https://doi.org/10.1016/j.fusengdes.2019.02.067},
issn = {0920-3796},
year = {2019},
date = {2019-09-01},
journal = {Fusion Engineering and Design},
volume = {146},
pages = {1321-1324},
abstract = {An Ion Cyclotron Range of Frequency (ICRF) system can provide power for a number of tasks, experimentally verified on present machines: heating and current drive, first wall conditioning, plasma startup, removing central impurities, controlling sawteeth and current ramp down assist. The system has a high plug-to-power efficiency and most of the components external to the machine are sturdy, with industrial steady state capability. Traditional ICRF antenna systems are often characterized by a high operating voltage and high power density. Low power density and low voltage however provides a bonus in terms of reliability. Therefore, travelling wave type antennas have been proposed (Ragona and Messiaen, 2016). They can be integrated in the blanket and use only a limited number of feeders. The effect on the tritium breeding ratio of such an antenna incorporated in the blanket, including the feeders, is small. The k// spectrum is peaked and the dominant k// value can be optimized for coupling and bulk absorption, while avoiding the generation of coaxial modes in the edge. The coupling can be further enhanced with gas puffing near the antenna. Assuming the ITER-2010-low density profile, 50 MW can be coupled with a voltage on the antenna components of about 15 kV.},
note = {SI:SOFT-30},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
An Ion Cyclotron Range of Frequency (ICRF) system can provide power for a number of tasks, experimentally verified on present machines: heating and current drive, first wall conditioning, plasma startup, removing central impurities, controlling sawteeth and current ramp down assist. The system has a high plug-to-power efficiency and most of the components external to the machine are sturdy, with industrial steady state capability. Traditional ICRF antenna systems are often characterized by a high operating voltage and high power density. Low power density and low voltage however provides a bonus in terms of reliability. Therefore, travelling wave type antennas have been proposed (Ragona and Messiaen, 2016). They can be integrated in the blanket and use only a limited number of feeders. The effect on the tritium breeding ratio of such an antenna incorporated in the blanket, including the feeders, is small. The k// spectrum is peaked and the dominant k// value can be optimized for coupling and bulk absorption, while avoiding the generation of coaxial modes in the edge. The coupling can be further enhanced with gas puffing near the antenna. Assuming the ITER-2010-low density profile, 50 MW can be coupled with a voltage on the antenna components of about 15 kV.
J. Ongena; A. Messiaen; A V Melnikov; R. Ragona; Y. O. Kazakov; D. Van Eester; Yu. N Dnestrovskii; P P Khvostenko; I N Roy; A N Romannikov
Conceptual study of an ICRH system for T-15MD using traveling wave antenna (TWA) sections Journal Article
In: Fusion Engineering and Design, vol. 146, pp. 787-791, 2019, ISSN: 0920-3796, (SI:SOFT-30).
@article{1935,
title = {Conceptual study of an ICRH system for T-15MD using traveling wave antenna (TWA) sections},
author = {J. Ongena and A. Messiaen and A V Melnikov and R. Ragona and Y. O. Kazakov and D. Van Eester and Yu. N Dnestrovskii and P P Khvostenko and I N Roy and A N Romannikov},
url = {http://www.sciencedirect.com/science/article/pii/S0920379619300882},
doi = {https://doi.org/10.1016/j.fusengdes.2019.01.080},
issn = {0920-3796},
year = {2019},
date = {2019-09-01},
journal = {Fusion Engineering and Design},
volume = {146},
pages = {787-791},
abstract = {This paper describes a conceptual ICRH system for the tokamak T-15MД, in construction at the Nuclear Fusion Institute of the Kurchatov Research Centre for Atomic Energy in Moscow. The proposed system consists of a number of Travelling Wave Antenna sections, located below the equatorial plane of the tokamak. This antenna system, loaded by a simulated density profile provided by the T-15MД team, is modelled including its resonant ring feeding system. Resonant ring feeding allows the recirculation of the RF power that is not radiated to the plasma and the termination of the TWA section on its iterative impedance. The paper describes the antenna design, the feeding ring tuning algorithm and expected performances of this antenna concept. The chosen geometry of the TWA sections is compatible with that of a future reactor and therefore this ICRH antenna system for T-15MД represents also a test bed for DEMO.},
note = {SI:SOFT-30},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
This paper describes a conceptual ICRH system for the tokamak T-15MД, in construction at the Nuclear Fusion Institute of the Kurchatov Research Centre for Atomic Energy in Moscow. The proposed system consists of a number of Travelling Wave Antenna sections, located below the equatorial plane of the tokamak. This antenna system, loaded by a simulated density profile provided by the T-15MД team, is modelled including its resonant ring feeding system. Resonant ring feeding allows the recirculation of the RF power that is not radiated to the plasma and the termination of the TWA section on its iterative impedance. The paper describes the antenna design, the feeding ring tuning algorithm and expected performances of this antenna concept. The chosen geometry of the TWA sections is compatible with that of a future reactor and therefore this ICRH antenna system for T-15MД represents also a test bed for DEMO.
Sunn T Pedersen; R K ö; M Jakubowski; M Krychowiak; D Gradic; C Killer; H Niemann; T Szepesi; U Wenzel; A Ali; G Anda; J Baldzuhn; T Barbui; C Biedermann; B D Blackwell; H -S Bosch; S Bozhenkov; R Brakel; S Brezinsek; J Cai; B Cannas; J W Coenen; J Cosfeld; A Dinklage; T Dittmar; P Drewelow; P Drews; D Dunai; F Effenberg; M Endler; Y Feng; J Fellinger; O Ford; H Frerichs; G Fuchert; Y Gao; J Geiger; A. Goriaev; K Hammond; J Harris; D Hathiramani; M Henkel; Y. O. Kazakov; A Kirschner; A Knieps; M Kobayashi; G Kocsis; P Kornejew; T Kremeyer; S Lazerzon; A LeViness; C Li; Y Li; Y Liang; S Liu; J Lore; S Masuzaki; V Moncada; O Neubauer; T T Ngo; J Oelmann; M Otte; V Perseo; F Pisano; Puig A Sitjes; M Rack; M Rasinski; J Romazanov; L Rudischhauser; G Schlisio; J C Schmitt; O Schmitz; B Schweer; S Sereda; M Sleczka; Y Suzuki; M Vecsei; E Wang; T. Wauters; S Wiesen; V Winters; G A Wurden; D Zhang; Zoletnik S and
First divertor physics studies in Wendelstein 7-X Journal Article
In: Nuclear Fusion, vol. 59, no. 9, pp. 096014, 2019.
@article{1933,
title = {First divertor physics studies in Wendelstein 7-X},
author = {Sunn T Pedersen and R K ö and M Jakubowski and M Krychowiak and D Gradic and C Killer and H Niemann and T Szepesi and U Wenzel and A Ali and G Anda and J Baldzuhn and T Barbui and C Biedermann and B D Blackwell and H -S Bosch and S Bozhenkov and R Brakel and S Brezinsek and J Cai and B Cannas and J W Coenen and J Cosfeld and A Dinklage and T Dittmar and P Drewelow and P Drews and D Dunai and F Effenberg and M Endler and Y Feng and J Fellinger and O Ford and H Frerichs and G Fuchert and Y Gao and J Geiger and A. Goriaev and K Hammond and J Harris and D Hathiramani and M Henkel and Y. O. Kazakov and A Kirschner and A Knieps and M Kobayashi and G Kocsis and P Kornejew and T Kremeyer and S Lazerzon and A LeViness and C Li and Y Li and Y Liang and S Liu and J Lore and S Masuzaki and V Moncada and O Neubauer and T T Ngo and J Oelmann and M Otte and V Perseo and F Pisano and Puig A Sitjes and M Rack and M Rasinski and J Romazanov and L Rudischhauser and G Schlisio and J C Schmitt and O Schmitz and B Schweer and S Sereda and M Sleczka and Y Suzuki and M Vecsei and E Wang and T. Wauters and S Wiesen and V Winters and G A Wurden and D Zhang and Zoletnik S and},
url = {https://doi.org/10.1088%2F1741-4326%2Fab280f},
doi = {10.1088/1741-4326/ab280f},
year = {2019},
date = {2019-09-01},
journal = {Nuclear Fusion},
volume = {59},
number = {9},
pages = {096014},
publisher = {IOP Publishing},
abstract = {The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment, which went into operation in 2015, has been operating since 2017 with an un-cooled modular graphite divertor. This allowed first divertor physics studies to be performed at pulse energies up to 80 MJ, as opposed to 4 MJ in the first operation phase, where five inboard limiters were installed instead of a divertor. This, and a number of other upgrades to the device capabilities, allowed extension into regimes of higher plasma density, heating power, and performance overall, e.g. setting a new stellarator world record triple product. The paper focuses on the first physics studies of how the island divertor works. The plasma heat loads arrive to a very high degree on the divertor plates, with only minor heat loads seen on other components, in particular baffle structures built in to aid neutral compression. The strike line shapes and locations change significantly from one magnetic configuration to another, in very much the same way that codes had predicted they would. Strike-line widths are as large as 10 cm, and the wetted areas also large, up to about 1.5 m2, which bodes well for future operation phases. Peak local heat loads onto the divertor were in general benign and project below the 10 MW m−2 limit of the future water-cooled divertor when operated with 10 MW of heating power, with the exception of low-density attached operation in the high-iota configuration. The most notable result was the complete (in all 10 divertor units) heat-flux detachment obtained at high-density operation in hydrogen.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment, which went into operation in 2015, has been operating since 2017 with an un-cooled modular graphite divertor. This allowed first divertor physics studies to be performed at pulse energies up to 80 MJ, as opposed to 4 MJ in the first operation phase, where five inboard limiters were installed instead of a divertor. This, and a number of other upgrades to the device capabilities, allowed extension into regimes of higher plasma density, heating power, and performance overall, e.g. setting a new stellarator world record triple product. The paper focuses on the first physics studies of how the island divertor works. The plasma heat loads arrive to a very high degree on the divertor plates, with only minor heat loads seen on other components, in particular baffle structures built in to aid neutral compression. The strike line shapes and locations change significantly from one magnetic configuration to another, in very much the same way that codes had predicted they would. Strike-line widths are as large as 10 cm, and the wetted areas also large, up to about 1.5 m2, which bodes well for future operation phases. Peak local heat loads onto the divertor were in general benign and project below the 10 MW m−2 limit of the future water-cooled divertor when operated with 10 MW of heating power, with the exception of low-density attached operation in the high-iota configuration. The most notable result was the complete (in all 10 divertor units) heat-flux detachment obtained at high-density operation in hydrogen.
R. Ragona; A. Messiaen; J M Bernard; E Delchambre; R Dumont; F. Durodié; J Hillairet; J. Ongena; D. Van Eester; M. Van Schoor
Traveling wave array for DEMO with proof of principle on WEST Journal Article
In: Fusion Engineering and Design, vol. 146, pp. 854-857, 2019, ISSN: 0920-3796, (SI:SOFT-30).
@article{1936,
title = {Traveling wave array for DEMO with proof of principle on WEST},
author = {R. Ragona and A. Messiaen and J M Bernard and E Delchambre and R Dumont and F. Durodié and J Hillairet and J. Ongena and D. Van Eester and M. Van Schoor},
url = {http://www.sciencedirect.com/science/article/pii/S0920379619301085},
doi = {https://doi.org/10.1016/j.fusengdes.2019.01.097},
issn = {0920-3796},
year = {2019},
date = {2019-09-01},
journal = {Fusion Engineering and Design},
volume = {146},
pages = {854-857},
abstract = {To decrease the power density and associated high voltages, a distributed antenna system is proposed as ICRH system for the DEMO reactor. Among the different solutions, a layout made from a set of travelling wave array (TWA) sections is considered as the most promising. It optimizes coupling to the plasma, is load resilient and avoids large values for the VSWR in the feeding lines. The total radiated power scales as the number of independently fed sections such that high reliability can be expected. The TWA concept for ICRH is innovative and very different from the traditional IC antennas. A test on WEST would provide a proof of principle of the validity of the TWA approach together with a comparison with the existing WEST IC antennas. The chosen geometry of the TWA section is compatible with one unit of the complete set designed for a future reactor. The paper describes the progress made in the preparation of a test on WEST along with the extrapolation for a future reactor like DEMO. A comparative modeling with the present antennas is also discussed and a preliminary RAMI analysis is introduced showing the promising positive impact of the TWA design on the RAMI scores.},
note = {SI:SOFT-30},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
To decrease the power density and associated high voltages, a distributed antenna system is proposed as ICRH system for the DEMO reactor. Among the different solutions, a layout made from a set of travelling wave array (TWA) sections is considered as the most promising. It optimizes coupling to the plasma, is load resilient and avoids large values for the VSWR in the feeding lines. The total radiated power scales as the number of independently fed sections such that high reliability can be expected. The TWA concept for ICRH is innovative and very different from the traditional IC antennas. A test on WEST would provide a proof of principle of the validity of the TWA approach together with a comparison with the existing WEST IC antennas. The chosen geometry of the TWA section is compatible with one unit of the complete set designed for a future reactor. The paper describes the progress made in the preparation of a test on WEST along with the extrapolation for a future reactor like DEMO. A comparative modeling with the present antennas is also discussed and a preliminary RAMI analysis is introduced showing the promising positive impact of the TWA design on the RAMI scores.
I. Shesterikov; K. Crombé; J. -M. Noterdaeme
Experimental analysis of the particle transport in the magnetized plasma column with an application to the helicon discharge Journal Article
In: Physics of Plasmas, vol. 26, no. 09, pp. 093508, 2019.
@article{1939,
title = {Experimental analysis of the particle transport in the magnetized plasma column with an application to the helicon discharge},
author = {I. Shesterikov and K. Crombé and J. -M. Noterdaeme},
year = {2019},
date = {2019-09-01},
journal = {Physics of Plasmas},
volume = {26},
number = {09},
pages = {093508},
abstract = {Different transport mechanisms in a magnetized radio frequency plasma discharge in the IShTAR device are compared. The total cross field particle transport systematically shows the best agreement with the turbulent diffusion. Also, the ion mobility dominated transport could substantially contribute to radial losses. The relative role of parallel and perpendicular losses in the overall particle confinement is also compared. The total perpendicular particle losses are comparable or even larger than the parallel ones, imposing a practical limitation on achieving high density plasma simply combining several helicon antennae sequentially arranged along the glass tube. For the same reason, reducing the diameter of a source tube, in an attempt to achieve a higher plasma density, could be not very efficient when perpendicular particle loss is dominant.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
Different transport mechanisms in a magnetized radio frequency plasma discharge in the IShTAR device are compared. The total cross field particle transport systematically shows the best agreement with the turbulent diffusion. Also, the ion mobility dominated transport could substantially contribute to radial losses. The relative role of parallel and perpendicular losses in the overall particle confinement is also compared. The total perpendicular particle losses are comparable or even larger than the parallel ones, imposing a practical limitation on achieving high density plasma simply combining several helicon antennae sequentially arranged along the glass tube. For the same reason, reducing the diameter of a source tube, in an attempt to achieve a higher plasma density, could be not very efficient when perpendicular particle loss is dominant.
J Baldzuhn; H Damm; C D Beidler; K McCarthy; N Panadero; C Biedermann; S A Bozhenkov; K J Brunner; G Fuchert; Y. O. Kazakov; M Beurskens; M Dibon; J Geiger; O Grulke; U H ö; T Klinger; F K ö; J Knauer; G Kocsis; P Kornejew; P T Lang; A Langenberg; H Laqua; N A Pablant; E Pasch; T S Pedersen; B Ploeckl; K Rahbarnia; G Schlisio; E R Scott; T Stange; A Stechow; T Szepesi; Y Turkin; F Wagner; V Winters; G Wurden; Zhang D and
Pellet fueling experiments in Wendelstein 7-X Journal Article
In: Plasma Physics and Controlled Fusion, vol. 61, no. 9, pp. 095012, 2019.
@article{1929,
title = {Pellet fueling experiments in Wendelstein 7-X},
author = {J Baldzuhn and H Damm and C D Beidler and K McCarthy and N Panadero and C Biedermann and S A Bozhenkov and K J Brunner and G Fuchert and Y. O. Kazakov and M Beurskens and M Dibon and J Geiger and O Grulke and U H ö and T Klinger and F K ö and J Knauer and G Kocsis and P Kornejew and P T Lang and A Langenberg and H Laqua and N A Pablant and E Pasch and T S Pedersen and B Ploeckl and K Rahbarnia and G Schlisio and E R Scott and T Stange and A Stechow and T Szepesi and Y Turkin and F Wagner and V Winters and G Wurden and Zhang D and},
url = {https://doi.org/10.1088%2F1361-6587%2Fab3567},
doi = {10.1088/1361-6587/ab3567},
year = {2019},
date = {2019-08-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {61},
number = {9},
pages = {095012},
publisher = {IOP Publishing},
abstract = {During the two most recent experimental campaigns in the advanced stellarator Wendelstein 7-X (W7-X) (Klinger et al 2017 Plasma Phys. Control. Fusion 59 014018; Bosch et al 2017 Nucl. Fusion 57 116015; Wolf et al 2017 Nucl. Fusion 57 102020; Pedersen et al 2017 Phys. Plasmas 24 0555030) hydrogen ice pellet injection was performed for the first time. In order to investigate the potential of pellet fueling in W7-X and to study the particle deposition in a large stellarator, a blower-gun system was installed with 40 pellets capability. The experience gained with this system will be used for the specification of a future steady-state pellet injector system. One important motivation for a pellet injector (Dibon 2014 Master-Thesis Technical University Munich, Max-Planck Institut IPP) on W7-X is the mitigation of hollow density profiles expected in case of predominant neoclassical transport. For long-pulse operation of up to 30 min, only electron cyclotron resonance heating is available on W7-X. Hence, pellet injection will be the only source for deep particle fueling. Deep particle fueling by pellets in tokamaks is supported by a grad-B drift, if the pellets are injected from the magnetic high-field-side. This approach was tested in W7-X, as well. The injection of series of pellets was also tested. Here, deep fueling is supported for later pellets in the series by the plasma cooling following the initial pellets in the same series. As in earlier experiments in the heliotron LHD (Takeiri et al 2017 Nucl. Fusion 57 102023), deep and rapid fueling could be achieved successfully in W7-X.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
During the two most recent experimental campaigns in the advanced stellarator Wendelstein 7-X (W7-X) (Klinger et al 2017 Plasma Phys. Control. Fusion 59 014018; Bosch et al 2017 Nucl. Fusion 57 116015; Wolf et al 2017 Nucl. Fusion 57 102020; Pedersen et al 2017 Phys. Plasmas 24 0555030) hydrogen ice pellet injection was performed for the first time. In order to investigate the potential of pellet fueling in W7-X and to study the particle deposition in a large stellarator, a blower-gun system was installed with 40 pellets capability. The experience gained with this system will be used for the specification of a future steady-state pellet injector system. One important motivation for a pellet injector (Dibon 2014 Master-Thesis Technical University Munich, Max-Planck Institut IPP) on W7-X is the mitigation of hollow density profiles expected in case of predominant neoclassical transport. For long-pulse operation of up to 30 min, only electron cyclotron resonance heating is available on W7-X. Hence, pellet injection will be the only source for deep particle fueling. Deep particle fueling by pellets in tokamaks is supported by a grad-B drift, if the pellets are injected from the magnetic high-field-side. This approach was tested in W7-X, as well. The injection of series of pellets was also tested. Here, deep fueling is supported for later pellets in the series by the plasma cooling following the initial pellets in the same series. As in earlier experiments in the heliotron LHD (Takeiri et al 2017 Nucl. Fusion 57 102023), deep and rapid fueling could be achieved successfully in W7-X.
A Czarnecka; N Krawczyk; P Jacquet; E. Lerche; V Bobkov; C Challis; D Frigione; J Graves; K D Lawson; M J Mantsinen; L Meneses; E Pawelec; T P ü; M Sertoli; M Valisa; D. Van Eester
In: Plasma Physics and Controlled Fusion, vol. 61, no. 8, pp. 085004, 2019.
@article{1927,
title = {Analysis of metallic impurity content by means of VUV and SXR diagnostics in hybrid discharges with hot-spots on the JET-ITER-like wall poloidal limiter},
author = {A Czarnecka and N Krawczyk and P Jacquet and E. Lerche and V Bobkov and C Challis and D Frigione and J Graves and K D Lawson and M J Mantsinen and L Meneses and E Pawelec and T P ü and M Sertoli and M Valisa and D. Van Eester},
url = {https://doi.org/10.1088%2F1361-6587%2Fab2100},
doi = {10.1088/1361-6587/ab2100},
year = {2019},
date = {2019-08-01},
journal = {Plasma Physics and Controlled Fusion},
volume = {61},
number = {8},
pages = {085004},
publisher = {IOP Publishing},
abstract = {In preparation for the upcoming JET D–T campaign, great effort has been devoted during the 2015–2016 JET campaigns with the ITER-like wall (ILW) to the extension of the high performance H-mode phase in baseline and hybrid scenarios. Hybrid discharges were the only ones that have been stopped by the real-time vessel protection system due hot-spot formation on the outboard poloidal limiter. Generation of hot-spots was linked to the application of high neutral beams injection and ion cyclotron resonance heating (ICRH) power. In tokamaks with high-Z plasma components, the use of ICRH heating is also accompanied by an increased metallic impurity content. Simultaneous control of hot-spot temperature and the core impurity content was crucial due to the fact that the same plasma-wall interaction mechanism is responsible for both phenomena. Impurity data collected by SXR, EUV and VUV diagnostics were able to provide for the first time comprehensive information concerning tungsten and mid-Z impurities such as nickel, iron, and cooper. To determine absolute mid-Z impurity concentrations a new relative calibration technique, compatible with JET-ILW, has been developed based on cross-calibration with a calibrated spectrometer via the quasicontinuum of W in the 200–400 Å wavelength range. In hybrid discharges, it was found that local D2 gas injection, plasma current, separatrix density, and fast ion losses appeared to impact hot-spot temperature and core impurity levels. Analysis showed a reduced maximum hot-spot temperature and impurity concentration at higher gas rate. Changes in the plasma current had a strong impact on the plasma-wall interaction, both via modifications in the edge density and in the fast ion losses. At constant gas injection rate, both the hot-spot temperature and the core impurity content decreased with the separatrix density. The main mechanism responsible for the formation of the hot-spots was found to be linked to the fast ion losses, but RF sheath effects may also be playing a role in the high limiter temperatures observed in these experiments.},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
In preparation for the upcoming JET D–T campaign, great effort has been devoted during the 2015–2016 JET campaigns with the ITER-like wall (ILW) to the extension of the high performance H-mode phase in baseline and hybrid scenarios. Hybrid discharges were the only ones that have been stopped by the real-time vessel protection system due hot-spot formation on the outboard poloidal limiter. Generation of hot-spots was linked to the application of high neutral beams injection and ion cyclotron resonance heating (ICRH) power. In tokamaks with high-Z plasma components, the use of ICRH heating is also accompanied by an increased metallic impurity content. Simultaneous control of hot-spot temperature and the core impurity content was crucial due to the fact that the same plasma-wall interaction mechanism is responsible for both phenomena. Impurity data collected by SXR, EUV and VUV diagnostics were able to provide for the first time comprehensive information concerning tungsten and mid-Z impurities such as nickel, iron, and cooper. To determine absolute mid-Z impurity concentrations a new relative calibration technique, compatible with JET-ILW, has been developed based on cross-calibration with a calibrated spectrometer via the quasicontinuum of W in the 200–400 Å wavelength range. In hybrid discharges, it was found that local D2 gas injection, plasma current, separatrix density, and fast ion losses appeared to impact hot-spot temperature and core impurity levels. Analysis showed a reduced maximum hot-spot temperature and impurity concentration at higher gas rate. Changes in the plasma current had a strong impact on the plasma-wall interaction, both via modifications in the edge density and in the fast ion losses. At constant gas injection rate, both the hot-spot temperature and the core impurity content decreased with the separatrix density. The main mechanism responsible for the formation of the hot-spots was found to be linked to the fast ion losses, but RF sheath effects may also be playing a role in the high limiter temperatures observed in these experiments.
I. Shesterikov; K. Crombé; A Kostic; D A Sitnikov; M Usoltceva; R Ochoukov; S Heuraux; J Moritz; E Faudot; F Fischer; H Faugel; H F ü; G Siegl; J. -M. Noterdaeme
IShTAR: A test facility to study the interaction between RF wave and edge plasmas Journal Article
In: Review of Scientific Instruments, vol. 90, no. 8, pp. 083506, 2019.
@article{1930,
title = {IShTAR: A test facility to study the interaction between RF wave and edge plasmas},
author = {I. Shesterikov and K. Crombé and A Kostic and D A Sitnikov and M Usoltceva and R Ochoukov and S Heuraux and J Moritz and E Faudot and F Fischer and H Faugel and H F ü and G Siegl and J. -M. Noterdaeme},
year = {2019},
date = {2019-08-01},
journal = {Review of Scientific Instruments},
volume = {90},
number = {8},
pages = {083506},
keywords = {},
pubstate = {published},
tppubtype = {article}
}
R C Wolf; A Alonso; S Akaslompolo; J Baldzhun; M Beurskens; C D Beidler; C Biedermann; H -S Bosch; S Bozhenkov; R Brakel; H Braune; S Brezinsek; K -J Brunner; H Damm; A Dinklage; P Drewelow; F Effenberg; Y Feng; O Ford; G Fuchert; Y Gao; J Geiger; O Grulke; N Harder; D Hartmann; P Helander; B Heinemann; M Hirsch; U H ö; C Hopf; K Ida; M Isobe; M W Jakubowski; Y. O. Kazakov; C Killer; T Klinger; J Knauer; R K ö; M Krychowiak; A Langenberg; H P Laqua; S Lazerson; P McNeely; S Marsen; N Marushchenko; R Nocentini; K Ogawa; G Orozco; M Osakabe; M Otte; D Pablant; E Pasch; A Pavone; M Porkolab; Puig A Sitjes; K Rahbarnia; R Riedl; N Rust; E Scott; H Schilling; R Schroeder; T Stange; Von A Stechow; E Strumberger; Sunn T Pedersen; J Svensson; H Thomson; Y Turkin; L Vano; T. Wauters; G Wurden; M Yoshinuma; M Zanini; D Zhang; Wendelstein 7-X Team
Performance of Wendelstein 7-X stellarator plasmas during the first divertor operation phase Journal Article
In: Physics of Plasmas, vol. 26, pp. 082504, 2019.
@article{1928,
title = {Performance of Wendelstein 7-X stellarator plasmas during the first divertor operation phase},
author = {R C Wolf and A Alonso and S Akaslompolo and J Baldzhun and M Beurskens and C D Beidler and C Biedermann and H -S Bosch and S Bozhenkov and R Brakel and H Braune and S Brezinsek and K -J Brunner and H Damm and A Dinklage and P Drewelow and F Effenberg and Y Feng and O Ford and G Fuchert and Y Gao and J Geiger and O Grulke and N Harder and D Hartmann and P Helander and B Heinemann and M Hirsch and U H ö and C Hopf and K Ida and M Isobe and M W Jakubowski and Y. O. Kazakov and C Killer and T Klinger and J Knauer and R K ö and M Krychowiak and A Langenberg and H P Laqua and S Lazerson and P McNeely and S Marsen and N Marushchenko and R Nocentini and K Ogawa and G Orozco and M Osakabe and M Otte and D Pablant and E Pasch and A Pavone and M Porkolab and Puig A Sitjes and K Rahbarnia and R Riedl and N Rust and E Scott and H Schilling and R Schroeder and T Stange and Von A Stechow and E Strumberger and Sunn T Pedersen and J Svensson and H Thomson and Y Turkin and L Vano and T. Wauters and G Wurden and M Yoshinuma and M Zanini and D Zhang and Wendelstein 7-X Team},
year = {2019},
date = {2019-08-01},
journal = {Physics of Plasmas},
volume = {26},
pages = {082504},
keywords = {},
pubstate = {published},
tppubtype = {article}
}