D. Borodin - Academia.edu (original) (raw)

Papers by D. Borodin

Research paper thumbnail of Surface roughness effect on Mo physical sputtering and re-deposition in the linear plasma device PSI-2 predicted by ERO2.0

Nuclear Materials and Energy

Research paper thumbnail of Material migration studies with an ITER first wall panel proxy on EAST

Nuclear Fusion

ABSTRACT The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges betwe... more ABSTRACT The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8 µm is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. Excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained.

Research paper thumbnail of ERO modelling of tungsten erosion in the linear plasma device PSI-2

Nuclear Materials and Energy

Series of experiments on tungsten (W) erosion and transport in Argon (Ar) plasma were conducted a... more Series of experiments on tungsten (W) erosion and transport in Argon (Ar) plasma were conducted at the linear plasma device PSI-2. W erosion was measured with three independent methods: WI spectroscopy, mass loss and quartz micro-balance (QMB) deposition sensor. Consistent set of data produced in these experiments was interpreted using the 3D ERO code simulations, which have reproduced all the main trends observed. Influence of the physical model assumptions (e.g. energy and angular distributions of sputtered particles) was demonstrated. The effect of WI effective quasi-metastable (MS) state population dynamics on spectroscopy measurements is shown; the characteristic relaxation time is determined. The measured physical sputtering yields for W are close to the simulated data obtained in the binary collision approximation (BCA) approach (SDTrimSP code). The remaining discrepancies between simulations and the experiment, mostly in spectroscopy, are accounted to the uncertainties in the plasma parameters and atomic data.

Research paper thumbnail of Modelling of deposition and erosion of injected WF6 and MoF6 in TEXTOR

Nuclear Materials and Energy

Tracer injection experiments in TEXTOR with MoF 6 and WF 6 lead to local deposition of about 6% f... more Tracer injection experiments in TEXTOR with MoF 6 and WF 6 lead to local deposition of about 6% for Mo and about 1% for W relative to the injected amount of Mo and W atoms. Modelling of these experiments has been done with ERO applying updated data for physical sputtering. The dissociation of the injected molecules has been treated in a simplified manner due to the lack of dissociation rate coefficients. However, with this it was possible to reproduce the observed radial penetration of Mo and W atoms into the plasma. The modelled local deposition efficiencies are about 50% for Mo and 60% for W assuming typical plasma parameters for the experimental conditions used. To reproduce the measured deposition efficiencies an enhancement factor for the erosion of deposited Mo and W has to be assumed (∼10 for Mo and ∼25 for W). Due to the rather low electron temperature T e of these plasma conditions (T e ∼15 eV at the location of injection), Mo and W are mostly sputtered by impurities whereas sputtering due to deuterium is negligible. A parameter study applying larger electron temperature leads to increased sputtering and thus to reduced local deposition efficiencies of about 30% for Mo and 5% for W. Though, even under these conditions enhanced erosion, albeit with reduced enhancement factors, is needed in the modelling to obtain the small measured deposition efficiencies.

Research paper thumbnail of First ERO2.0 modeling of Be erosion and non-local transport in JET ITER-like wall

Physica Scripta

ERO is a Monte-Carlo code for modeling plasma-wall interaction and 3D plasma impurity transport f... more ERO is a Monte-Carlo code for modeling plasma-wall interaction and 3D plasma impurity transport for applications in fusion research. The code has undergone a significant upgrade (ERO2.0) which allows increasing the simulation volume in order to cover the entire plasma edge of a fusion device, allowing a more self-consistent treatment of impurity transport and comparison with a larger number and variety of experimental diagnostics. In this contribution, the physics-relevant technical innovations of the new code version are described and discussed. The new capabilities of the code are demonstrated by modeling of beryllium (Be) erosion of the main wall during JET limiter discharges. Results for erosion patterns along the limiter surfaces and global Be transport including incident particle distributions are presented. A novel synthetic diagnostic, which mimics experimental wide-angle 2D camera images, is presented and used for validating various aspects of the code, including erosion, magnetic shadowing, non-local impurity transport, and light emission simulation.

Research paper thumbnail of Modelling of 13CH4 injection through graphite and tungsten test limiters in the scrape-off layer of TEXTOR using the coupled ERO-SDTrimSP code

Research paper thumbnail of Modeling of erosion and deposition of ITER limiters during ramp phases

Research paper thumbnail of An analytical model for the electric field and particle tracing at surface vicinity in plasma-surface interaction experiments

Research paper thumbnail of Study of physical and chemical sputtering of Beryllium in the JET ITER-Like Wall

Research paper thumbnail of Modelling of local carbon deposition from methane and ethene injection through graphite and tungsten test limiters in TEXTOR

Plasma Physics and Controlled Fusion, 2010

Research paper thumbnail of Close coupling approach for heavy particle collisions with an excited atom: transitions between n =3 states in He

Physica Scripta, 2006

The close coupling (CC) equations in the impact parameter (IP) representation are considered. Cal... more The close coupling (CC) equations in the impact parameter (IP) representation are considered. Calculations for transitions in the 3s-3p-3d He atom system induced by collisions with single charged ions were carried out. Results are compared with those obtained in the frame of the Born approach. The effects of stepped transitions, normalization, channel coupling and their dependence on interaction strength are discussed.

Research paper thumbnail of Estimations of erosion fluxes, material deposition and tritium retention in the divertor of ITER

Journal of Nuclear Materials, 2009

Material mixing effects on erosion/deposition and tritium retention in ITER have been modelled wi... more Material mixing effects on erosion/deposition and tritium retention in ITER have been modelled with ERO. It is seen that target lifetime is much less critical than tritium retention. A long-term tritium retention rate of ~9mg T/s is obtained, assuming constant beryllium concentration in the background plasma of 0.1% for outer and 1% for inner divertor. Retention in inner is about twice that in outer divertor. Using a TriDyn-based model for mixed layers instead of homogenous material mixing does change the detailed profiles of erosion and redeposition. However, overall tritium retention is almost unchanged. Also, profiles for the beryllium influx along the divertor plates calculated with DivImp have been used for the ERO modelling. The resulting beryllium flux on the targets decreases by factors of 25 (inner) and 55 (outer divertor) compared to constant concentrations used so far. First ERO calculations using these beryllium profiles indicate a reduction of overall tritium retention by a factor of ~4 (compared to constant beryllium concentrations as used before), mainly due to reduced beryllium deposition.

Research paper thumbnail of Penetration depths of injected/sputtered tungsten in the plasma edge layer of TEXTOR

Journal of Nuclear Materials, 2013

Research paper thumbnail of Investigations of castellated structures for ITER: The effect of castellation shaping and alignment on fuel retention and impurity deposition in gaps

Journal of Nuclear Materials, 2009

Castellation will be used in divertor and first wall components to provide thermomechanical stabi... more Castellation will be used in divertor and first wall components to provide thermomechanical stability of ITER. Radioactive fuel may be stored in the gaps of castellated structures representing a safety issue for ITER. Tungsten castellated structures with different shapes were exposed in TEXTOR to investigate the impact of cell shaping on impurity transport and fuel deposition in the gaps. 2 After exposure a significant intermixing of tungsten was detected in carbon deposits in the gaps reaching 70 at. % W in the deposition layer. This will provide difficulties in cleaning the gaps in ITER. Poloidal gaps of shaped cells contained a factor or 3 less deuterium than those of rectangular cells, the carbon deposition exhibited only marginal advantages of a new geometry. Poloidal and toroidal gaps contained comparable amount of C and D. Significant deposition at the bottom of gaps was measured which could only partly be reproduced by modeling.

Research paper thumbnail of Modelling of impurity deposition in gaps of castellated surfaces with the 3D-GAPS code

Plasma Physics and Controlled Fusion, 2010

The Monte-Carlo neutral transport code 3D-GAPS is described. The code models impurity transport a... more The Monte-Carlo neutral transport code 3D-GAPS is described. The code models impurity transport and deposition in remote areas, such as gaps between cells of castellated plasma-facing surfaces. A step-by-step investigation of the interplay of different processes that may influence the deposition inside gaps, namely particle reflection, elastic neutral collisions, different particle sources, chemical erosion and plasma penetration into gaps, is presented. Examples of modelling results in application to the TEXTOR experiment with a castellated test limiter are provided. It is shown that only with the assumption of the presence of species with different reflection probabilities, do simulated carbon deposition profiles agree with experimental observations for side surfaces of the gaps. These species can be attributed to different particle sources, e.g. carbon atoms and hydrocarbon radicals. Background carbon ions and atoms have low and moderate values of the reflection coefficient (R &l...

Research paper thumbnail of Spectra of O II in the plasma boundary of TEXTOR-94

Plasma Physics and Controlled Fusion, 2002

Spectra of the O II ion were obtained with temporal and spatial (along the poloidal radius) resol... more Spectra of the O II ion were obtained with temporal and spatial (along the poloidal radius) resolution by injecting of oxygen-containing molecules through a block of the upper poloidal limiter of TEXTOR-94 into the boundary plasma. Three different wavelength regions (4660±100, 4340±100, 3730±100 Å) were observed during reproducible TEXTOR-94 discharges with n e = 1×10 18 m −3 , T e = 90 eV at the last closed flux surface (LCFS). For calibration purposes the oxygen emission was simultaneously registered in all discharges with a 2D camera using the light from an O II line at 4416 Å. Radial intensity distributions of lines corresponding to the transitions 2p 2 [ 3 P]3s-2p 2 [ 3 P]3p (components of the multiplets 4 P-4 S, 4 P-4 P, 4 P-4 D, 2 P-2 D) and 2p 2 [ 1 D]3s-2p 2 [ 1 D]3p (doublet 2 D-2 D) were measured. It is shown that the relative intensities of the lines inside the multiplets are in good agreement with theoretical data calculated in SL-coupling. An appropriate collisional-radiative model (the level list and the corresponding atomic database) has been developed and calculations have been carried out using the GKU kinetic code developed at the P.N. Lebedev Institute. Cross-sections of atomic processes have been calculated using the ATOM code by the K-matrix method. A comparison of theoretical and experimental data is presented. It is shown that the relative multiplet intensities have a weak dependence on electron density and temperature. The populations of the ground configuration states of O II were investigated. Absolute values for the 'ionization per photon' were measured and a comparison with modelled values is given.

Research paper thumbnail of RF sheath-enhanced beryllium sources at JET’s ICRH antennas

ABSTRACT Local beryllium (Be) I and Be II line intensities were measured in the plasma-wall inter... more ABSTRACT Local beryllium (Be) I and Be II line intensities were measured in the plasma-wall interaction region near an ICRH antenna in JET. The intent was to use these intensities as a measure of the formation of local Radio Frequency (RF) sheath potentials, through RF sheath rectification and potential build up at the end of field lines passing in front of the antenna. Experimentally, it was found that the Be I and Be II emission increase when using the antenna local to the spectroscopic measurement, and increase even more when using a remote antenna that is magnetically connected to the observation point. Magnetic field mapping indicates a magnetic connection between the observation location and the top corner region of the remote antenna and/or its protection limiter. These measurements can be used in support of RF sheath modeling that is an important part of the optimization of antenna design for next generation fusion energy devices, including ITER.

Research paper thumbnail of Plasma Waves: Basic Physics, Absorption, Emission, Heating, Current-Drive

Research paper thumbnail of Spectroscopical observation of Si

Plasma Physics and Controlled Fusion

ABSTRACT

Research paper thumbnail of Beryllium migration in JET ITER-like wall plasmas

Nuclear Fusion, 2015

ABSTRACT JET is used as a test bed for ITER, to investigate beryllium migration which connects th... more ABSTRACT JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (Ein = 35 eV) and more than 100%, caused by Be self-sputtering (Ein = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at Ein = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.

Research paper thumbnail of Surface roughness effect on Mo physical sputtering and re-deposition in the linear plasma device PSI-2 predicted by ERO2.0

Nuclear Materials and Energy

Research paper thumbnail of Material migration studies with an ITER first wall panel proxy on EAST

Nuclear Fusion

ABSTRACT The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges betwe... more ABSTRACT The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8 µm is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. Excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained.

Research paper thumbnail of ERO modelling of tungsten erosion in the linear plasma device PSI-2

Nuclear Materials and Energy

Series of experiments on tungsten (W) erosion and transport in Argon (Ar) plasma were conducted a... more Series of experiments on tungsten (W) erosion and transport in Argon (Ar) plasma were conducted at the linear plasma device PSI-2. W erosion was measured with three independent methods: WI spectroscopy, mass loss and quartz micro-balance (QMB) deposition sensor. Consistent set of data produced in these experiments was interpreted using the 3D ERO code simulations, which have reproduced all the main trends observed. Influence of the physical model assumptions (e.g. energy and angular distributions of sputtered particles) was demonstrated. The effect of WI effective quasi-metastable (MS) state population dynamics on spectroscopy measurements is shown; the characteristic relaxation time is determined. The measured physical sputtering yields for W are close to the simulated data obtained in the binary collision approximation (BCA) approach (SDTrimSP code). The remaining discrepancies between simulations and the experiment, mostly in spectroscopy, are accounted to the uncertainties in the plasma parameters and atomic data.

Research paper thumbnail of Modelling of deposition and erosion of injected WF6 and MoF6 in TEXTOR

Nuclear Materials and Energy

Tracer injection experiments in TEXTOR with MoF 6 and WF 6 lead to local deposition of about 6% f... more Tracer injection experiments in TEXTOR with MoF 6 and WF 6 lead to local deposition of about 6% for Mo and about 1% for W relative to the injected amount of Mo and W atoms. Modelling of these experiments has been done with ERO applying updated data for physical sputtering. The dissociation of the injected molecules has been treated in a simplified manner due to the lack of dissociation rate coefficients. However, with this it was possible to reproduce the observed radial penetration of Mo and W atoms into the plasma. The modelled local deposition efficiencies are about 50% for Mo and 60% for W assuming typical plasma parameters for the experimental conditions used. To reproduce the measured deposition efficiencies an enhancement factor for the erosion of deposited Mo and W has to be assumed (∼10 for Mo and ∼25 for W). Due to the rather low electron temperature T e of these plasma conditions (T e ∼15 eV at the location of injection), Mo and W are mostly sputtered by impurities whereas sputtering due to deuterium is negligible. A parameter study applying larger electron temperature leads to increased sputtering and thus to reduced local deposition efficiencies of about 30% for Mo and 5% for W. Though, even under these conditions enhanced erosion, albeit with reduced enhancement factors, is needed in the modelling to obtain the small measured deposition efficiencies.

Research paper thumbnail of First ERO2.0 modeling of Be erosion and non-local transport in JET ITER-like wall

Physica Scripta

ERO is a Monte-Carlo code for modeling plasma-wall interaction and 3D plasma impurity transport f... more ERO is a Monte-Carlo code for modeling plasma-wall interaction and 3D plasma impurity transport for applications in fusion research. The code has undergone a significant upgrade (ERO2.0) which allows increasing the simulation volume in order to cover the entire plasma edge of a fusion device, allowing a more self-consistent treatment of impurity transport and comparison with a larger number and variety of experimental diagnostics. In this contribution, the physics-relevant technical innovations of the new code version are described and discussed. The new capabilities of the code are demonstrated by modeling of beryllium (Be) erosion of the main wall during JET limiter discharges. Results for erosion patterns along the limiter surfaces and global Be transport including incident particle distributions are presented. A novel synthetic diagnostic, which mimics experimental wide-angle 2D camera images, is presented and used for validating various aspects of the code, including erosion, magnetic shadowing, non-local impurity transport, and light emission simulation.

Research paper thumbnail of Modelling of 13CH4 injection through graphite and tungsten test limiters in the scrape-off layer of TEXTOR using the coupled ERO-SDTrimSP code

Research paper thumbnail of Modeling of erosion and deposition of ITER limiters during ramp phases

Research paper thumbnail of An analytical model for the electric field and particle tracing at surface vicinity in plasma-surface interaction experiments

Research paper thumbnail of Study of physical and chemical sputtering of Beryllium in the JET ITER-Like Wall

Research paper thumbnail of Modelling of local carbon deposition from methane and ethene injection through graphite and tungsten test limiters in TEXTOR

Plasma Physics and Controlled Fusion, 2010

Research paper thumbnail of Close coupling approach for heavy particle collisions with an excited atom: transitions between n =3 states in He

Physica Scripta, 2006

The close coupling (CC) equations in the impact parameter (IP) representation are considered. Cal... more The close coupling (CC) equations in the impact parameter (IP) representation are considered. Calculations for transitions in the 3s-3p-3d He atom system induced by collisions with single charged ions were carried out. Results are compared with those obtained in the frame of the Born approach. The effects of stepped transitions, normalization, channel coupling and their dependence on interaction strength are discussed.

Research paper thumbnail of Estimations of erosion fluxes, material deposition and tritium retention in the divertor of ITER

Journal of Nuclear Materials, 2009

Material mixing effects on erosion/deposition and tritium retention in ITER have been modelled wi... more Material mixing effects on erosion/deposition and tritium retention in ITER have been modelled with ERO. It is seen that target lifetime is much less critical than tritium retention. A long-term tritium retention rate of ~9mg T/s is obtained, assuming constant beryllium concentration in the background plasma of 0.1% for outer and 1% for inner divertor. Retention in inner is about twice that in outer divertor. Using a TriDyn-based model for mixed layers instead of homogenous material mixing does change the detailed profiles of erosion and redeposition. However, overall tritium retention is almost unchanged. Also, profiles for the beryllium influx along the divertor plates calculated with DivImp have been used for the ERO modelling. The resulting beryllium flux on the targets decreases by factors of 25 (inner) and 55 (outer divertor) compared to constant concentrations used so far. First ERO calculations using these beryllium profiles indicate a reduction of overall tritium retention by a factor of ~4 (compared to constant beryllium concentrations as used before), mainly due to reduced beryllium deposition.

Research paper thumbnail of Penetration depths of injected/sputtered tungsten in the plasma edge layer of TEXTOR

Journal of Nuclear Materials, 2013

Research paper thumbnail of Investigations of castellated structures for ITER: The effect of castellation shaping and alignment on fuel retention and impurity deposition in gaps

Journal of Nuclear Materials, 2009

Castellation will be used in divertor and first wall components to provide thermomechanical stabi... more Castellation will be used in divertor and first wall components to provide thermomechanical stability of ITER. Radioactive fuel may be stored in the gaps of castellated structures representing a safety issue for ITER. Tungsten castellated structures with different shapes were exposed in TEXTOR to investigate the impact of cell shaping on impurity transport and fuel deposition in the gaps. 2 After exposure a significant intermixing of tungsten was detected in carbon deposits in the gaps reaching 70 at. % W in the deposition layer. This will provide difficulties in cleaning the gaps in ITER. Poloidal gaps of shaped cells contained a factor or 3 less deuterium than those of rectangular cells, the carbon deposition exhibited only marginal advantages of a new geometry. Poloidal and toroidal gaps contained comparable amount of C and D. Significant deposition at the bottom of gaps was measured which could only partly be reproduced by modeling.

Research paper thumbnail of Modelling of impurity deposition in gaps of castellated surfaces with the 3D-GAPS code

Plasma Physics and Controlled Fusion, 2010

The Monte-Carlo neutral transport code 3D-GAPS is described. The code models impurity transport a... more The Monte-Carlo neutral transport code 3D-GAPS is described. The code models impurity transport and deposition in remote areas, such as gaps between cells of castellated plasma-facing surfaces. A step-by-step investigation of the interplay of different processes that may influence the deposition inside gaps, namely particle reflection, elastic neutral collisions, different particle sources, chemical erosion and plasma penetration into gaps, is presented. Examples of modelling results in application to the TEXTOR experiment with a castellated test limiter are provided. It is shown that only with the assumption of the presence of species with different reflection probabilities, do simulated carbon deposition profiles agree with experimental observations for side surfaces of the gaps. These species can be attributed to different particle sources, e.g. carbon atoms and hydrocarbon radicals. Background carbon ions and atoms have low and moderate values of the reflection coefficient (R &l...

Research paper thumbnail of Spectra of O II in the plasma boundary of TEXTOR-94

Plasma Physics and Controlled Fusion, 2002

Spectra of the O II ion were obtained with temporal and spatial (along the poloidal radius) resol... more Spectra of the O II ion were obtained with temporal and spatial (along the poloidal radius) resolution by injecting of oxygen-containing molecules through a block of the upper poloidal limiter of TEXTOR-94 into the boundary plasma. Three different wavelength regions (4660±100, 4340±100, 3730±100 Å) were observed during reproducible TEXTOR-94 discharges with n e = 1×10 18 m −3 , T e = 90 eV at the last closed flux surface (LCFS). For calibration purposes the oxygen emission was simultaneously registered in all discharges with a 2D camera using the light from an O II line at 4416 Å. Radial intensity distributions of lines corresponding to the transitions 2p 2 [ 3 P]3s-2p 2 [ 3 P]3p (components of the multiplets 4 P-4 S, 4 P-4 P, 4 P-4 D, 2 P-2 D) and 2p 2 [ 1 D]3s-2p 2 [ 1 D]3p (doublet 2 D-2 D) were measured. It is shown that the relative intensities of the lines inside the multiplets are in good agreement with theoretical data calculated in SL-coupling. An appropriate collisional-radiative model (the level list and the corresponding atomic database) has been developed and calculations have been carried out using the GKU kinetic code developed at the P.N. Lebedev Institute. Cross-sections of atomic processes have been calculated using the ATOM code by the K-matrix method. A comparison of theoretical and experimental data is presented. It is shown that the relative multiplet intensities have a weak dependence on electron density and temperature. The populations of the ground configuration states of O II were investigated. Absolute values for the 'ionization per photon' were measured and a comparison with modelled values is given.

Research paper thumbnail of RF sheath-enhanced beryllium sources at JET’s ICRH antennas

ABSTRACT Local beryllium (Be) I and Be II line intensities were measured in the plasma-wall inter... more ABSTRACT Local beryllium (Be) I and Be II line intensities were measured in the plasma-wall interaction region near an ICRH antenna in JET. The intent was to use these intensities as a measure of the formation of local Radio Frequency (RF) sheath potentials, through RF sheath rectification and potential build up at the end of field lines passing in front of the antenna. Experimentally, it was found that the Be I and Be II emission increase when using the antenna local to the spectroscopic measurement, and increase even more when using a remote antenna that is magnetically connected to the observation point. Magnetic field mapping indicates a magnetic connection between the observation location and the top corner region of the remote antenna and/or its protection limiter. These measurements can be used in support of RF sheath modeling that is an important part of the optimization of antenna design for next generation fusion energy devices, including ITER.

Research paper thumbnail of Plasma Waves: Basic Physics, Absorption, Emission, Heating, Current-Drive

Research paper thumbnail of Spectroscopical observation of Si

Plasma Physics and Controlled Fusion

ABSTRACT

Research paper thumbnail of Beryllium migration in JET ITER-like wall plasmas

Nuclear Fusion, 2015

ABSTRACT JET is used as a test bed for ITER, to investigate beryllium migration which connects th... more ABSTRACT JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (Ein = 35 eV) and more than 100%, caused by Be self-sputtering (Ein = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at Ein = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.