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Hellsten, Torbjörn
Publications (10 of 378) Show all publications
Moradi, S., Rachlew, E., Bergsåker, H., Frassinetti, L., Garcia Carrasco, A., Hellsten, T., . . . et al., . (2020). Global scaling of the heat transport in fusion plasmas. Physical Review Research, 2
Open this publication in new window or tab >>Global scaling of the heat transport in fusion plasmas
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2020 (English)In: Physical Review Research, E-ISSN 2643-1564, Vol. 2Article in journal (Refereed) Published
Abstract [en]

A global heat flux model based on a fractional derivative of plasma pressure is proposed for the heat transport in fusion plasmas. The degree of the fractional derivative of the heat flux, α, is defined through the power balance analysis of the steady state. The model was used to obtain the experimental values of α for a large database of the Joint European Torus (JET) carbon-wall as well as ITER like-wall plasmas. The fractional degrees of the electron heat flux are found to be α<2, for all the selected pulses in the database, suggesting a deviation from the diffusive paradigm. Moreover, the results show that as the volume integrated input power is increased, the fractional degree of the electron heat flux converges to α∼0.8, indicating a global scaling between the net heating and the pressure profile in the high-power JET plasmas. The model is expected to provide insight into the proper kinetic description for the fusion plasmas and improve the accuracy of the heat transport predictions.

National Category
Medical Laboratory Technologies
Identifiers
urn:nbn:se:kth:diva-314094 (URN)10.1103/PhysRevResearch.2.013027 (DOI)000600701000006 ()2-s2.0-85085553415 (Scopus ID)
Note

QC 20220615

Available from: 2022-06-15 Created: 2022-06-15 Last updated: 2025-02-09Bibliographically approved
Vallejos, P., Jonsson, T., Ragona, R., Van Eester, D., Zaar, B. & Hellsten, T. (2020). Iterative addition of finite Larmor radius effects to finite element models using wavelet decomposition. Plasma Physics and Controlled Fusion, 62(4), Article ID 045022.
Open this publication in new window or tab >>Iterative addition of finite Larmor radius effects to finite element models using wavelet decomposition
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2020 (English)In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 62, no 4, article id 045022Article in journal (Refereed) Published
Abstract [en]

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.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2020
Keywords
Morlet wavelets, finite element method, ion cyclotron resonance heating, mode conversion, ion Bernstein waves
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-271924 (URN)10.1088/1361-6587/ab6f55 (DOI)000521361100001 ()2-s2.0-85086036895 (Scopus ID)
Note

QC 20200421

Available from: 2020-04-21 Created: 2020-04-21 Last updated: 2024-12-21Bibliographically approved
Zanca, P., Bergsåker, H., Bykov, I., Frassinetti, L., Garcia Carrasco, A., Hellsten, T., . . . et al, . (2019). A power-balance model of the density limit in fusion plasmas: application to the L-mode tokamak. Nuclear Fusion, 59(12), Article ID 126011.
Open this publication in new window or tab >>A power-balance model of the density limit in fusion plasmas: application to the L-mode tokamak
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2019 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, no 12, article id 126011Article in journal (Refereed) Published
Abstract [en]

A power-balance model, with radiation losses from impurities and neutrals, gives a unified description of the density limit (DL) of the stellarator, the L-mode tokamak, and the reversed field pinch (RFP). The model predicts a Sudo-like scaling for the stellarator, a Greenwald- like scaling, alpha I-p(8/9), for the RFP and the ohmic tokamak, a mixed scaling, alpha (PIp4/9)-I-4/9, for the additionally heated L-mode tokamak. In a previous paper (Zanca et al 2017 Nucl. Fusion 57 056010) the model was compared with ohmic tokamak, RFP and stellarator experiments. Here, we address the issue of the DL dependence on heating power in the L-mode tokamak. Experimental data from high-density disrupted L-mode discharges performed at JET, as well as in other machines, arc taken as a term of comparison. The model fits the observed maximum densities better than the pure Greenwald limit.

Place, publisher, year, edition, pages
Institute of Physics Publishing (IOPP), 2019
Keywords
magnetohydrodynamics, transport, radiation
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-269131 (URN)10.1088/1741-4326/ab3b31 (DOI)000488059900001 ()2-s2.0-85076758927 (Scopus ID)
Note

QC 20200312

Available from: 2020-03-12 Created: 2020-03-12 Last updated: 2024-03-15Bibliographically approved
Pamela, S., Bergsåker, H., Bykov, I., Frassinetti, L., Garcia Carrasco, A., Hellsten, T., . . . et al, . (2019). A wall-aligned grid generator for non-linear simulations of MHD instabilities in tokamak plasmas. Computer Physics Communications, 243, 41-50
Open this publication in new window or tab >>A wall-aligned grid generator for non-linear simulations of MHD instabilities in tokamak plasmas
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2019 (English)In: Computer Physics Communications, ISSN 0010-4655, E-ISSN 1879-2944, Vol. 243, p. 41-50Article in journal (Refereed) Published
Abstract [en]

Block-structured mesh generation techniques have been well addressed in the CFD community for automobile and aerospace studies, and their applicability to magnetic fusion is highly relevant, due to the complexity of the plasma-facing wall structures inside a tokamak device. Typically applied to non-linear simulations of MHD instabilities relevant to magnetically confined fusion, the JOREK code was originally developed with a 2D grid composed of isoparametric bi-cubic Bezier finite elements, that are aligned to the magnetic equilibrium of tokamak plasmas (the third dimension being represented by Fourier harmonics). To improve the applicability of these simulations, the grid-generator has been generalised to provide a robust extension method, using a block-structured mesh approach, which allows the simulations of arbitrary domains of tokamak vacuum vessels. Such boundary-aligned grids require the adaptation of boundary conditions along the edge of the new domain. Demonstrative non-linear simulations of plasma edge instabilities are presented to validate the robustness of the new grid, and future potential physics applications for tokamak plasmas are discussed. The methods presented here may be of interest to the wider community, beyond tokamak physics, wherever imposing arbitrary boundaries to quadrilateral finite elements is required.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Fusion, Tokamak, MHD, Instability, ELM, Grid
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-269148 (URN)10.1016/j.cpc.2019.05.007 (DOI)000474316900005 ()2-s2.0-85066828087 (Scopus ID)
Note

QC 20200311

Available from: 2020-03-11 Created: 2020-03-11 Last updated: 2024-03-15Bibliographically approved
Henderson, S. S., Bergsåker, H., Bykov, I., Frassinetti, L., Garcia Carrasco, A., Hellsten, T., . . . et al., . (2019). An assessment of nitrogen concentrations from spectroscopic measurements in the JET and ASDEX upgrade divertor. Nuclear Materials and Energy, 18, 147-152
Open this publication in new window or tab >>An assessment of nitrogen concentrations from spectroscopic measurements in the JET and ASDEX upgrade divertor
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2019 (English)In: Nuclear Materials and Energy, E-ISSN 2352-1791, Vol. 18, p. 147-152Article in journal (Refereed) Published
Abstract [en]

The impurity concentration in the tokamak divertor plasma is a necessary input for predictive scaling of divertor detachment, however direct measurements from existing tokamaks in different divertor plasma conditions are limited. To address this, we have applied a recently developed spectroscopic N II line ratio technique for measuring the N concentration in the divertor to a range of H-mode and L-mode plasma from the ASDEX Upgrade and JET tokamaks, respectively. The results from both devices show that as the power crossing the separatrix, P-sep, is increased under otherwise similar core conditions (e.g. density), a higher N concentration is required to achieve the same detachment state. For example, the N concentrations at the start of detachment increase from approximate to 2% to approximate to 9% as P-sep, is increased from approximate to 2.5 MW to approximate to 7 MW. These results tentatively agree with scaling law predictions (e.g. Goldston et al.) motivating a further study examining the parameters which affect the N concentration required to reach detachment. Finally, the N concentrations from spectroscopy and the ratio of D and N gas valve fluxes agree within experimental uncertainty only when the vessel surfaces are fully-loaded with N.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Impurity, Nitrogen, Divertor, Concentration, Spectroscopy, Tokamak
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-270861 (URN)10.1016/j.nme.2018.12.012 (DOI)000460107500026 ()2-s2.0-85058630263 (Scopus ID)
Note

QC 20200316

Available from: 2020-03-16 Created: 2020-03-16 Last updated: 2024-03-15Bibliographically approved
Ström, P., Petersson, P., Rubel, M., Bergsåker, H., Bykov, I., Frassinetti, L., . . . et al., . (2019). Analysis of deposited layers with deuterium and impurity elements on samples from the divertor of JET with ITER-like wall. Journal of Nuclear Materials, 516, 202-213
Open this publication in new window or tab >>Analysis of deposited layers with deuterium and impurity elements on samples from the divertor of JET with ITER-like wall
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2019 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 516, p. 202-213Article in journal (Refereed) Published
Abstract [en]

Inconel-600 blocks and stainless steel covers for quartz microbalance crystals from remote corners in the JET-ILW divertor were studied with time-of-flight elastic recoil detection analysis and nuclear reaction analysis to obtain information about the areal densities and depth profiles of elements present in deposited material layers. Surface morphology and the composition of dust particles were examined with scanning electron microscopy and energy-dispersive X-ray spectroscopy. The analyzed components were present in JET during three ITER-like wall campaigns between 2010 and 2017. Deposited layers had a stratified structure, primarily made up of beryllium, carbon and oxygen with varying atomic fractions of deuterium, up to more than 20%. The range of carbon transport from the ribs of the divertor carrier was limited to a few centimeters, and carbon/deuterium co-deposition was indicated on the Inconel blocks. High atomic fractions of deuterium were also found in almost carbon-free layers on the quartz microbalance covers. Layer thicknesses up to more than 1 micrometer were indicated, but typical values were on the order of a few hundred nanometers. Chromium, iron and nickel fractions were less than or around 1% at layer surfaces while increasing close to the layer-substrate interface. The tungsten fraction depended on the proximity of the plasma strike point to the divertor corners. Particles of tungsten, molybdenum and copper with sizes less than or around 1 micrometer were found. Nitrogen, argon and neon were present after plasma edge cooling and disruption mitigation. Oxygen-18 was found on component surfaces after injection, indicating in-vessel oxidation. Compensation of elastic recoil detection data for detection efficiency and ion-induced release of deuterium during the measurement gave quantitative agreement with nuclear reaction analysis, which strengthens the validity of the results.

Keywords
Fusion, Tokamak, Plasma-wall interactions, ToF-ERDA, NRA, SEM
National Category
Fusion, Plasma and Space Physics
Research subject
Physics
Identifiers
urn:nbn:se:kth:diva-240616 (URN)10.1016/j.jnucmat.2018.11.027 (DOI)000458897100020 ()2-s2.0-85060313456 (Scopus ID)
Note

QC 20190125

Available from: 2018-12-20 Created: 2018-12-20 Last updated: 2022-09-05Bibliographically approved
Drenik, A., Bergsåker, H., Bykov, I., Frassinetti, L., Garcia Carrasco, A., Hellsten, T., . . . Zychor, I. (2019). Analysis of the outer divertor hot spot activity in the protection video camera recordings at JET. Fusion engineering and design, 139, 115-123
Open this publication in new window or tab >>Analysis of the outer divertor hot spot activity in the protection video camera recordings at JET
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2019 (English)In: Fusion engineering and design, ISSN 0920-3796, E-ISSN 1873-7196, Vol. 139, p. 115-123Article in journal (Refereed) Published
Abstract [en]

Hot spots on the divertor tiles at JET result in overestimation of the tile surface temperature which causes unnecessary termination of pulses. However, the appearance of hot spots can also indicate the condition of the divertor tile surfaces. To analyse the behaviour of the hot spots in the outer divertor tiles of JET, a simple image processing algorithm is developed. The algorithm isolates areas of bright pixels in the camera image and compares them to previously identified hot spots. The activity of the hot spots is then linked to values of other signals and parameters in the same time intervals. The operation of the detection algorithm was studied in a limited pulse range with high hot spot activity on the divertor tiles 5, 6 and 7. This allowed us to optimise the values of the controlling parameters. Then, the wider applicability of the method has been demonstrated by the analysis of the hot spot behaviour in a whole experimental campaign.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
JET, ITER-like wall, Plasma-wall interaction, Image analysis
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-269599 (URN)10.1016/j.fusengdes.2018.12.079 (DOI)000458939100016 ()2-s2.0-85059687937 (Scopus ID)
Note

QC 20200407

Available from: 2020-04-07 Created: 2020-04-07 Last updated: 2022-12-12Bibliographically approved
Orsitto, F. P., Bergsåker, H., Bykov, I., Frassinetti, L., Garcia-Carrasco, A., Hellsten, T., . . . et al., . (2019). Approximate analytic expressions using Stokes model for tokamak polarimetry and their range of validity. Plasma Physics and Controlled Fusion, 61(5), Article ID 055008.
Open this publication in new window or tab >>Approximate analytic expressions using Stokes model for tokamak polarimetry and their range of validity
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2019 (English)In: Plasma Physics and Controlled Fusion, ISSN 0741-3335, E-ISSN 1361-6587, Vol. 61, no 5, article id 055008Article in journal (Refereed) Published
Abstract [en]

The analysis of the polarimetry measurements has the aim of validating models (De Marco and Segre 1972 Plasma Phys. 14 245), with a careful attention to the clarification of their limits of application. In this paper a new approximation method is introduced, the so-called special constant Omega direction (SCOD), which gives an analytical solution to the polarimetry exact Stokes model equations. The available approximate solutions (including SCOD) of the polarimetry propagation equations are presented, compared and their application limits determined, using a reference tokamak configuration, which is a simplified equilibrium for a circular tokamak. The SCOD approximation is compared successfully to the Stokes model in the context also of equilibria evaluated for two JET discharges. The approximation methods are analytical or very simple mathematical expressions which can also be used in equilibrium codes for their optimization.

Place, publisher, year, edition, pages
IOP PUBLISHING LTD, 2019
Keywords
plasma diagnostics, polarimetry, equilibrium reconstruction
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-270513 (URN)10.1088/1361-6587/ab09c2 (DOI)000462886500001 ()2-s2.0-85069514831 (Scopus ID)
Note

QC 20200416

Available from: 2020-04-16 Created: 2020-04-16 Last updated: 2024-03-15Bibliographically approved
Romazanov, J., Bergsåker, H., Bykov, I., Frassinetti, L., Garcia Carrasco, A., Hellsten, T., . . . et al., . (2019). Beryllium global erosion and deposition at JET-ILW simulated with ERO2.0. Nuclear Materials and Energy, 18, 331-338
Open this publication in new window or tab >>Beryllium global erosion and deposition at JET-ILW simulated with ERO2.0
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2019 (English)In: Nuclear Materials and Energy, E-ISSN 2352-1791, Vol. 18, p. 331-338Article in journal (Refereed) Published
Abstract [en]

The recently developed Monte-Carlo code ERO2.0 is applied to the modelling of limited and diverted discharges at JET with the ITER-like wall (ILW). The global beryllium (Be) erosion and deposition is simulated and compared to experimental results from passive spectroscopy. For the limiter configuration, it is demonstrated that Be self-sputtering is an important contributor (at least 35%) to the Be erosion. Taking this contribution into account, the ERO2.0 modelling confirms previous evidence that high deuterium (D) surface concentrations of up to similar to 50% atomic fraction provide a reasonable estimate of Be erosion in plasma-wetted areas. For the divertor configuration, it is shown that drifts can have a high impact on the scrape-off layer plasma flows, which in turn affect global Be transport by entrainment and lead to increased migration into the inner divertor. The modelling of the effective erosion yield for different operational phases (ohmic, L- and H-mode) agrees with experimental values within a factor of two, and confirms that the effective erosion yield decreases with increasing heating power and confinement.

Place, publisher, year, edition, pages
Elsevier, 2019
Keywords
Beryllium, Erosion, ER02.0, JET ITER-like wall
National Category
Fusion, Plasma and Space Physics
Identifiers
urn:nbn:se:kth:diva-270863 (URN)10.1016/j.nme.2019.01.015 (DOI)000460107500056 ()2-s2.0-85061047660 (Scopus ID)
Note

QC 20200316

Available from: 2020-03-16 Created: 2020-03-16 Last updated: 2022-06-26Bibliographically approved
Telesca, G., Bergsåker, H., Bykov, I., Frassinetti, L., Fridström, R., Garcia Carrasco, A., . . . et al., . (2019). COREDIV numerical simulation of high neutron rate JET-ILW DD pulses in view of extension to JET-ILW DT experiments. Nuclear Fusion, 59(5), Article ID 056026.
Open this publication in new window or tab >>COREDIV numerical simulation of high neutron rate JET-ILW DD pulses in view of extension to JET-ILW DT experiments
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2019 (English)In: Nuclear Fusion, ISSN 0029-5515, E-ISSN 1741-4326, Vol. 59, no 5, article id 056026Article in journal (Refereed) Published
Abstract [en]

Two high performance JET-ILW pulses, pertaining to the 2016 experimental campaign, have been numerically simulated with the self-consistent code COREDIV with the aim of predicting the ELM-averaged power load to the target when extrapolated to DT plasmas. The input power of about 33 MW as well as the total radiated power and the average density are similar in the two pulses, but for one of them the density is provided by combined low gas puff and pellet injection, characterized by low SOL density, for the other one by gas fuelling only, at higher SOT. density. Considering the magnetic configuration of theses pulses and the presence of a significant amount of Ni (not included in the version of the code used for these simulations), a number of assumptions are made in order to reproduce numerically the main core and SOL experimental data. The extrapolation to DT plasmas at the original input power of 33 MW, and taking into account only the thermal component of the alpha-power, does not show any significant difference regarding the power to the target with respect to the DD case. In contrast, the simulations at auxiliary power 40 MW, both at the original I-p = 3 MA and at I-p = 4 MA, show that the power to the target for both pulses is possibly too high to be sustained for about 5 s by strike-point sweeping alone without any control by Ne seeding. Even though the target power load may decrease to about 13-15 MW with substantial Ne seeding for both pulses, as from numerical predictions, there are indications suggesting that the control of the power load may be more critical for the pulse with pellet injection, due to the reduced SOL radiation.

Place, publisher, year, edition, pages
Institute of Physics Publishing (IOPP), 2019
Keywords
tokamak, integrated modeling, neon seeding, JET-ILW
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-270847 (URN)10.1088/1741-4326/ab0c47 (DOI)000464453100002 ()2-s2.0-85066072535 (Scopus ID)
Note

QC 20200317

Available from: 2020-03-17 Created: 2020-03-17 Last updated: 2024-03-15Bibliographically approved
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