D. Borodin
Forschungszentrum Jülich
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Featured researches published by D. Borodin.
Nuclear Fusion | 2015
S. Brezinsek; A. Widdowson; M. Mayer; V. Philipps; P. Baron-Wiechec; J. W. Coenen; K. Heinola; A. Huber; J. Likonen; Per Petersson; M. Rubel; M. Stamp; D. Borodin; J.P. Coad; A.G. Carrasco; A. Kirschner; S. Krat; K. Krieger; B. Lipschultz; Ch. Linsmeier; G. F. Matthews; K. Schmid; Jet Contributors
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.
Plasma Physics and Controlled Fusion | 2010
D. Matveev; A. Kirschner; A. Litnovsky; M. Komm; D. Borodin; V. Philipps; G. Van Oost
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 ≤ 0.6), while some of the hydrocarbon radicals produced by chemical erosion of redeposited carbon layers have high reflection probability (R ≥ 0.9). Deposition at the bottom of the gaps cannot be adequately reproduced unless extreme assumptions on particle sources and reflection properties are imposed. Elastic neutral collisions and ionization of neutrals escaping the gaps have no significant influence on the results. Nevertheless, particle-in-cell simulations of plasma penetration into gaps are essential for estimating the incoming ion flux and leading to a better quantitative agreement with experimental observations.
Physica Scripta | 2011
D. Borodin; A. Kirschner; S. Carpentier-Chouchana; R.A. Pitts; S. Lisgo; C. Björkas; P.C. Stangeby; J.D. Elder; A Galonska; D. Matveev; V. Philipps; U. Samm
Previous studies (Carpentier et al 2011 J. Nucl. Mater. 415 S165–S169) carried out with the LIM code of the ITER first wall (FW) on beryllium (Be) erosion, re-deposition and tritium retention by co-deposition under steady-state burning plasma conditions have shown that, depending on input plasma parameter assumptions and sputtering yields, the erosion lifetime and fuel retention on some parts of the FW can be a serious concern. The importance of the issue is such that a benchmark of this previous work is sought and has been provided by the ERO code (Pitts et al 2011 J. Nucl. Mater. 415 S957–S964) simulations described in this paper. Provided that inputs to the codes are carefully matched, excellent agreement is found between the erosion/deposition profiles from both codes for a given ITER-shaped FW panel. Issues regarding the difficult problem of the correct treatment of Be sputtering are discussed in relation to the simulations. The possible influence of intrinsic Be impurity is investigated.
Plasma Physics and Controlled Fusion | 2008
S. Droste; A. Kirschner; D. Borodin; A. Kreter; S. Brezinsek; V. Philipps; U. Samm; O. Schmitz
The 3D Monte-Carlo code ERO, which calculates erosion processes, impurity transport and deposition, has been coupled to the Monte-Carlo code SDTrimSP to simulate material mixing processes in wall components more precisely. SDTrimSP calculates the transport of ions in solids by means of the binary collision approximation. It keeps track of the depth dependent material concentration caused by implantation of projectiles in the solid. Modelling with the coupled code ERO-SDTrimSP is compared with dedicated TEXTOR experiments, in which the formation of mixed surface layers has been studied. In these experiments, methane 13CH4 was injected through graphite and tungsten spherical limiters during plasma exposure and the local redeposition probability was measured post mortem by surface analysis. A significant difference in the carbon 13C deposition efficiency, i.e. the ratio of the locally deposited to the injected amount of 13C, between graphite and tungsten was found, 4% for graphite and 0.3% for tungsten. Modelling of these experiments with ERO-SDTrimSP reproduces the clear substrate dependence with about 2% deposition efficiency on graphite and less than 0.5% on tungsten in good agreement with the experiment. The reason for the substrate dependence is partly explained by the higher physical sputtering yield of a thin carbon film on top of a tungsten substrate compared with a graphite substrate. Surface roughness of the materials has been identified to be another important parameter for the interpretation of the results.
Nuclear Fusion | 2014
S. Brezinsek; M. Stamp; D. Nishijima; D. Borodin; S. Devaux; K. Krieger; S. Marsen; M. O'Mullane; C. Bjoerkas; A. Kirschner; Jet-Efda Contributors
The effective sputtering yield of Be was determined in situ by emission spectroscopy of low ionizing Be as function of the deuteron impact energy (Ein = 25–175 eV) and Be surface temperature (Tsurf = 200 °C–520 °C) in limiter discharges carried out in the JET tokamak. Be self sputtering dominates the erosion at high impact energies (Ein > 150 eV) and causes far beyond 1. drops to low values, below 4.5%, at the accessible lowest impact energy (Ein 25 eV) achievable in limiter configuration. At medium impact energies, Ein = 75 eV, two contributors to the measured of 9% were identified: two third of the eroded Be originates from bare physical sputtering and one third from chemical assisted physical sputtering . The later mechanism has been clearly identified by the appearance of BeD A–X emission and quantified in cause of a temperature dependence at which the BeD practically vanishes at highest observed Be limiter temperatures. The recorded Tsurf dependence, obtained in a series of 34 identical discharges with ratch-up of Tsurf by plasma impact and inertial cooling after the discharge, revealed that the reduction of BeD is correlated with an increase of D2 emission. The release mechanism of deuterium in the Be interaction layer is exchanged under otherwise constant recycling flux conditions at the limiter.The reduction of with Tsurf is also correlated to the reduction of the Be content in the core plasma providing information on the total source strength and Be screening. The chemical assisted physical sputtering, always present at the nominal limiter pre-heating temperature of Tsurf = 200 °C, is associated with an additional sputtering channel with respect to ordinary physical sputtering which is surface temperature independent. These JET experiments in limiter configuration are used to benchmark the ERO code and verify ITER first wall erosion prediction. The ERO code overestimates the observed Be sputtering in JET by a factor of about 2.5 which can be transferred to ITER predictions and prolong the expected lifetime of first wall elements.
Journal of Physics B | 2010
A. Pospieszczyk; D. Borodin; S. Brezinsek; A. Huber; A. Kirschner; Ph. Mertens; G. Sergienko; B. Schweer; I. L. Beigman; L Vainshtein
Rate coefficients for the excitation and ionization of neutral as well as singly ionized particles and-–predominantly-–their ratios S(D)/XB, which are important for the conversion of photon into particle fluxes in ionizing fusion boundary plasma, have been modelled and experimentally determined in boundary plasmas for fusion-relevant species such as He I, Li I, C I, B I&II, O I&II, Si I&II, Mo I, W I&II, H2, CH(D), C2.
Plasma Physics and Controlled Fusion | 2013
C. Björkas; D. Borodin; A. Kirschner; R. K. Janev; D. Nishijima; R.P. Doerner; K. Nordlund
The low-energy erosion mechanisms of first-wall materials subject to a fusion plasma are poorly known theoretically, even though they pose a critical problem for the development of tokamak-like fusion reactors. Using molecular dynamics computer simulations and analytical theory, we have examined the fundamental mechanisms of the erosion of first-wall materials, focusing on molecular release from beryllium surfaces. We show that the observed sputtering of BeD molecules from beryllium when exposed to a D plasma can be explained by the swift chemical sputtering mechanism, and that it also can happen in BeW alloys. This demonstrates that pure metals can, in contrast to conventional wisdom, be sputtered chemically. We also link the simulations of BeD sputtering to the plasma impurity transport code ERO, in order to follow the behavior of sputtered BeD species in a plasma. This multi-scale approach enables direct comparisons with experimental observations of BeD sputtering in the PISCES-B facility.
Plasma Physics and Controlled Fusion | 2013
Shuyu Dai; A. Kirschner; D. Matveev; D. Borodin; C Björkas; Jizhong Sun; Dezhen Wang
A Monte-Carlo code called SURO was developed to study the influence of surface roughness on the impurity deposition characteristics in fusion experiments. SURO uses the test particle approach to describe the impact of background plasma and the deposition of impurity particles on a sinusoidal surface. The local impact angle and dynamic change of surface roughness as well as surface concentrations of different species due to erosion and deposition are taken into account. Coupled with the three-dimensional Monte-Carlo code ERO, SURO was used to study the impact of surface roughness on 13C deposition in 13CH4 injection experiments in TEXTOR. The simulations showed that the amount of net deposited 13C species increased with surface roughness. Parameter studies with varying 12C and 13C fluxes were performed to gain insights into impurity deposition characteristics on the rough surface. Calculations of the exposure time needed for surface smoothing for TEXTOR and ITER were also carried out for different scenarios.
Plasma Physics and Controlled Fusion | 2012
van Ga Gijs Swaaij; K. Bystrov; D. Borodin; A. Kirschner; van der Lb Vegt; van Gj Gerard Rooij; G. De Temmerman; Wj Goedheer
For understanding carbon erosion and redeposition in nuclear fusion devices, it is important to understand the transport and chemical break-up of hydrocarbon molecules in edge plasmas, often diagnosed by emission of the CH A 2Δ–X 2Π Gero band around 430 nm. The CH A-level can be excited either by electron-impact (EI) or by dissociative recombination (DR) of hydrocarbon ions. These processes were included in the 3D Monte Carlo impurity transport code ERO. A series of methane injection experiments was performed in the high-density, low-temperature linear plasma generator Pilot-PSI, and simulated emission intensity profiles were benchmarked against these experiments. It was confirmed that excitation by DR dominates at Te < 1.5 eV. The results indicate that the fraction of DR events that lead to a CH radical in the A-level and consequent photon emission is at least 10%. Additionally, quenching of the excited CH radicals by EI de-excitation was included in the modeling. This quenching is shown to be significant: depending on the electron density, it reduces the effective CH emission by a factor of 1.4 at ne = 1.3 × 1020 m−3, to 2.8 at ne = 9.3 × 1020 m−3. Its inclusion significantly improved agreement between experiment and modeling.
Physica Scripta | 2011
S. Brezinsek; D. Borodin; J. W. Coenen; D Kondratjew; M. Laengner; A. Pospieszczyk; U. Samm
Tungsten is foreseen as the plasma-facing component material for baffles, the dome and strike-point area in the ITER divertor. Quantification of the W source, which is connected with the components lifetime and W plasma concentration, is one of the outstanding issues in the qualification process. A dedicated experiment in TEXTOR with the exposure of a W/C twin limiter to the near scrape-off layer plasma has been carried out in order to address the W sputtering and local material mixing in the electron temperature range between Te=30 and 85?eV, achieved with deuterium fueling in four steps. The Te range is comparable to the baffle region and the strike-point area during non-detached transient phases of the ITER divertor plasma. Quantification of the W sputtering yield and the impinging impurity fluxes was performed with the aid of optical spectroscopy, in particular by observation of WI and WII lines. As no inverse photon efficiencies in the plasma parameter range of the twin limiter experiment exist, we performed in a second experiment for the first time a calibration of WI and WII photon efficiencies with local injection of WF6 through a gas inlet into the TEXTOR edge plasma. The in situ determined effective inverse photon efficiency of about 85 for the most prominent WI line at 400.9?nm, which is in good agreement with GKU modelling for the covered Te range and 650 for the WII line at 434.8?nm, has been applied to the corresponding photon fluxes in the twin limiter experiment. The W sputtering yield decreases from 5.2 to 0.5%, thus by about one order of magnitude, with a reduction of Te from 85?eV down to 30?eV and a simultaneous increase of the impinging deuterium ion flux by 50% occurs. A lower limit for the prompt redeposition has been estimated at 50% by analyzing the WI to WII flux ratio. Local measurement of OII (441.6?nm) and CII (426.7?nm) provided impurity flux ratios of 0.6% for O and 5.2% for C related to the deuterium recycling, respectively ion flux. Both flux ratios remain constant for all phases of the discharge with plasma edge cooling. W erosion is predominantly caused by sputtering of higher ionization stages of O and C impinging on the W limiter half and not by the fuel species itself. Plasma cooling below the physical sputtering threshold could not be achieved without impurity seeding.