Robert V Budny | Princeton University (original) (raw)

Papers by Robert V Budny

Plasma Physics and Controlled Fusion, 1999

... This is generally observed when the following operating conditions are satisfied simultaneous... more ... This is generally observed when the following operating conditions are satisfied simultaneously: (i) radiative power fraction Prad/Ptot ⩾ 50%; (ii) sufficiently high central chord averaged electron density ¯ne0 such that the Greenwald number ¯ne0/¯ne0,Gr exceeds 0.70, (where ...

Long period sawteeth have been observed to result in the low-β triggering of neo-classical tearin... more Long period sawteeth have been observed to result in the low-β triggering of neo-classical tearing modes (NTMs), which can significantly degrade plasma confinement. In ITER, the stabilising effects of the fusion-born α particles are likely to exacerbate this. Consequently, it is important to develop a detailed physical understanding of sawtooth behaviour. Recent work on plasmas heated using neutral beam injection (NBI) in JET, MAST, ASDEX Upgrade and TEXTOR has provided significant insight into the physical processes that determine sawtooth behaviour. The asymmetric dependence of sawtooth period upon neutral beam power injected in either the co-or counter-current direction exhibited in both JET and TEXTOR can be explained by including the effects of the passing energetic ions, the toroidal flow and the flow shear. In JET, the counter-passing fast ions destabilise the n = 1 internal kink mode-thought to be associated with sawteeth-whilst the flow shear strongly influences the stabilisation from the trapped particles. The sawtooth behaviour in TEXTOR has been explained through a combination of both kinetic effects and stabilisation from toroidal rotation. In order to avoid triggering NTMs, many techniques have been proposed to control, and in particular, to destabilise the sawtooth oscillations. Here, sawtooth behaviour in off-axis NBI-heated plasmas in JET, MAST and ASDEX Upgrade is presented. It is found that the energetic particles born outside the q = 1 surface due to the off-axis NBI can destabilise the sawteeth, even in the presence of stabilising on-axis fast particles. We also consider sawtooth control using ion cyclotron resonance heating (ICRH). Previously it had been assumed that the change in sawtooth behaviour in ICRH heated plasmas arose due to the change in the magnetic shear at q = 1. The energetic passing ions are found to influence the internal kink mode when the distribution of ions is asymmetric in v , a natural feature of co or counter propagating ICRH waves.

An assessment of the 2.45 and 14 MeV neutron yields and neutron emission spatial distributions fo... more An assessment of the 2.45 and 14 MeV neutron yields and neutron emission spatial distributions for the DD and DT phases of ITER operation has been carried out by self-consistent 1.5D transport simulations of the core plasma with ASTRA. This assessment is needed to define the details of a set of dedicated experiments to be carried out in order to establish the reliability of the neutronics calculations early in ITER nuclear operations [1] and to evaluate the capabilities and performance of the ITER neutron diagnostics in different phases/scenarios during ITER nuclear operation. The main challenge for these two issues is that the evaluations of the capability/reliability are foreseen to be carried out during the ITER DD operational phase (or early in the DT phase) when the neutron flux and fluence will be orders of magnitude lower than those expected at Q = 10. Our assessment includes Ohmic, Lmode and H-mode plasmas with full (5.3T) and half (2.65T) toroidal field. The separatrix boun...

Nuclear Fusion, 2018

This paper further analyses the EDGE2D-EIRENE simulations presented by [Chankin et al 2017 Nucl. ... more This paper further analyses the EDGE2D-EIRENE simulations presented by [Chankin et al 2017 Nucl. Mater. Energy 12 273], of L-mode JET plasmas in vertical-vertical (VV) and vertical-horizontal (VH) divertor configurations. As expected, the simulated outer divertor ionisation source peaks near the separatrix in VV and radially further out in VH. We identify the reflections of recycled neutrals from lower divertor tiles as the primary mechanism by which ionisation is concentrated on the outer divertor separatrix in the VV configuration. These lower tile reflection pathways (of neutrals from the outer divertor, and to an even greater extent from the inner divertor) dominate the outer divertor separatrix ionisation. In contrast, the lower-tile-reflection pathways are much weaker in the VH simulation and its outer divertor ionisation is dominated by neutrals which do not reflect from any surfaces. Interestingly, these differences in neutral pathways give rise to strong differences in the heat flux density width λq at the outer divertor entrance: λq = 3.2 mm in VH compared to λq = 11.8 mm in VV. In VH, a narrow channel exists in the near SOL where the convected heat flux, driven by strong Er × B flow and thermoelectric current, dominates over the conducted heat flux. The width of this channel sets λq and is determined by the radial distance between the separatrix and the ionisation peak in the outer divertor.

None. Consists of a list of team members and affiliations.

Plasma Physics and Controlled Fusion, 2002

Perturbative transport experiments have been performed in JET low or reverse magnetic shear plasm... more Perturbative transport experiments have been performed in JET low or reverse magnetic shear plasmas either in conditions of fully developed internal transport barrier (ITB) or during a phase where an ITB was not observed. Transient peripheral cooling was induced either by laser ablation or by shallow pellet injection, and the ensuing travelling cold pulse was used to probe the plasma

Nuclear Fusion, 2003

In JET advanced tokamak research mainly focuses on plasmas with internal transport barriers (ITBs... more In JET advanced tokamak research mainly focuses on plasmas with internal transport barriers (ITBs), generated by modifications of the current profile. The formerly developed optimised shear regime with low magnetic shear in the plasma center has been extended to deeply reversed magnetic shear configurations. High fusion performance with wide ITBs has been obtained transiently in deeply reversed magnetic shear configuration: HIPB98(y,2)~1.9, βN=2.4 at Ip=2.5MA. At somewhat reduced performance electron and ion ITBs have been sustained in full current drive operation with 1MA of bootstrap current: HIPB98(y,2)~1, βN=1.7 at Ip=2.0MA. The ITBs have been maintained up to 11s. This duration, much larger than the energy confinement time (37 times larger), is already approaching a current resistive time. New real-time measurements and feedback control algorithms have been developed and implemented in JET for successfully controlling the ITB dynamics and the current density profile in the highly non-inductive regime.

Fusion Engineering and Design, 2019

Nuclear Materials and Energy, 2019

Recent achievements in studies of the effects of ICRF (Ion Cyclotron Range of Frequencies) power ... more Recent achievements in studies of the effects of ICRF (Ion Cyclotron Range of Frequencies) power on the SOL (Scrape-Off Layer) and PWI (Plasma Wall Interactions) in ASDEX Upgrade (AUG), Alcator C-Mod, and JET-ILW are reviewed. Capabilities to diagnose and model the effect of DC biasing and associated impurity production at active antennas and on magnetic field connections to antennas are described. The experiments show that ICRF near-fields can lead not only to E × B convection, but also to modifications of the SOL density, which for Alcator C-Mod are limited to a narrow region near antenna. On the other hand, the SOL density distribution along with impurity sources can be tailored using local gas injection in AUG and JET-ILW with a positive effect on reduction of impurity sources. The technique of RF image current cancellation at antenna limiters was successfully applied in AUG using the 3-strap AUG antenna and extended to the 4-strap Alcator C-Mod field-aligned antenna. Multiple observations confirmed the reduction of the impact of ICRF on the SOL and on total impurity production when the ratio of the power of the central straps to the total antenna power is in the range 0.6 < P cen / P total < 0.8. Near-field calculations indicate that this fairly robust technique can be applied to the ITER ICRF antenna, enabling the mode of operation with reduced PWI. On the contrary, for the A2 antenna in JET-ILW the technique is hindered by RF sheaths excited at the antenna septum. Thus, in order to reduce the effect of ICRF power on PWI in a future fusion reactor, the antenna design has to be optimized along with design of plasmafacing components.

Supershot profiles were used to simulate plasmas in a neutralbeam-driven tokamak reactor designed... more Supershot profiles were used to simulate plasmas in a neutralbeam-driven tokamak reactor designed to achieve fusion energy production with Q = 2-3. Profiles from a TFTR supershot were scaled to larger radii, density, and electron temperature. The TRANSP code was used to calculate performance of these plasmas. Examples are given of steady-state plasmas with large beam-driven bootstrap cur rents. The required energy transport rate is comparable to that in TFTR. but the particle transport rate must be less. The PEST code indicates that the plasmas would be MHD stable if the central q can be controlled.

Tb« report was prepared as an account of v^ork sponsored by an agency of Cfic United States Oover... more Tb« report was prepared as an account of v^ork sponsored by an agency of Cfic United States OovernmenL Neither the United States Government nor any agency thereof, nor any of their _ ^^ (*% employees, makes any warranly, exp<icss or implied, or assumes any legal liability or responsi-^ ^F E^ L£ bility for the accuracy, completeness, or usefulness of any information, apparatus, product, or & A /V ^fc| 1 ^^ |^» process disclosed, or represents that its use would not irifringe privately owned rights. Refer-M\W\MT^**^ ence herein to any specific commercial product, process, or service by trade name, trademark. Jl » *^ manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommenda:ion, or favoring ty (he Untied Stales Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily stare or reflect I hose of the United States Government or any agency thereof.

AIP Conference Proceedings, 1995

A peak fusion power production of 9.3±0.7 MW has been achieved on the Tokamak Fusion Test Reactor... more A peak fusion power production of 9.3±0.7 MW has been achieved on the Tokamak Fusion Test Reactor (TFTR) in deuterium plasmas heated by co and counter injected deuterium and tritium neutral beams with a total power of 33.7 MW. The ratio of fusion power output to ...

Physical Review D, 1974

ABSTRACT

Routine operation in the enhanced energy confinement (or H-mode) reqime during neutral beara inje... more Routine operation in the enhanced energy confinement (or H-mode) reqime during neutral beara injection was achieved by modifying the pnx divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to T E /I p) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (Auyoj < 0,1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, th; edge relaxation phenomena tERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikis in the divertor plasma density, H a emission, and on the X-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high P T in the H-mode discharges were hampered bv a deterioration in the H-mode confinement and major disruptions which limited the achievable p" T. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical P boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n =1/1 mode with no observable external precursor oscillations.

Regimes of High Confinement (H-mode) have been studied in the Alcator C-Mod tokamak [Hutchinson e... more Regimes of High Confinement (H-mode) have been studied in the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Plasmas with no Edge Localized Modes (ELM-free) have been compared in detail to a new regime, Enhanced Dα (EDA). EDA discharges have only slightly lower energy confinement than comparable ELM-free ones, but show markedly reduced impurity confinement. Thus EDA discharges do not accumulate impurities and typically have a lower fraction of radiated power. The edge gradients in EDA seem to be relaxed by a continuous process rather than an intermittent one as is the case for standard ELMy discharges and thus do not present the first wall with large periodic heat loads. This process is probably related to fluctuations seen in the plasma edge. EDA plasmas are more likely at low plasma current (q > 3.7), for moderate plasma shaping, (δ 0 .35-0.55), and for high neutral pressures. As observed in soft x-ray emission, the pedestal width is found to scale with the same parameters that determine the EDA/ELM-free boundary.

Physical Review Letters, 1995

The toroidicity-induced Alfven eigenmodes(TAE) arefound to be stable in the Tokamak Fusion Test R... more The toroidicity-induced Alfven eigenmodes(TAE) arefound to be stable in the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium plasmas. The dominant stabilizing mechanisms are beam ion Landau damping and radiative damping. A core localized TAF ...

Nuclear Fusion, 2007

This paper reports on the recent studies of toroidal and poloidal momentum transport in JET. The ... more This paper reports on the recent studies of toroidal and poloidal momentum transport in JET. The ratio of the global energy confinement time to the momentum confinement is found to be close to τ E /τ φ =1 except for the low density discharges where the ratio is τ E /τ φ =2-3. On the other hand, local transport analysis of tens of discharges shows that the ratio of the local effective momentum diffusivity to the ion heat diffusivity is χ φ /χ i 0.1-0.4 rather than unity, as expected from the global confinement times and used in ITER predictions. The apparent discrepancy in the global and local momentum versus ion heat transport is explained by the fact that momentum confinement within edge pedestal is worse than that of the ion heat and thus, momentum pedestal is weaker than that of ion temperature. Another observation is that while the T i has a threshold in R/L Ti and profiles are stiff, the gradient in v φ increases with increasing torque and no threshold is found. Predictive transport simulations also confirm that χ φ /χ i 0.1-0.4 reproduce the core toroidal velocity profiles well. Concerning poloidal velocities on JET, the experimental measurements show that the carbon poloidal velocity can be an order of magnitude above the neo-classical estimate within the ITB. This significantly affects the calculated radial electric field and therefore, the E×B flow shear used for example in transport simulations. The Weiland model reproduces the onset, location and strength of the ITB well when the experimental poloidal rotation is used while it does not predict an ITB using the neo-classical poloidal velocity. The most plausible explanation for the generation of the anomalous poloidal velocity is the turbulence driven flow through the Reynold's stress. Both TRB and CUTIE turbulence codes show the existence of an anomalous poloidal velocity, being significantly larger than the neo-classical values. And similarly to experiments, the poloidal velocity profiles peak in the vicinity of the ITB and is caused by flow due to the Reynold's stress.

Nuclear Fusion, 2011

The experiment described in this paper is aimed at characterizing of ELMy H-mode for plasmas with... more The experiment described in this paper is aimed at characterizing of ELMy H-mode for plasmas with varying momentum input, rotation, power deposition profiles and ion to electron heating ratio obtained by varying the proportion between IC and NB heating. The motivation for the experiment was to verify that the basic confinement and transport properties of the baseline ITER Hmode are robust to these changes, and similar to those derived from the large database of NB heated H-modes. No significant difference in the density and temperature profiles or in the global confinement were found. Although ion temperature profiles were seen to be globally stiff, some variation of stiffness was obtained in the experiment by varying the deposition profiles, but not one that could significantly affect the profiles in terms of global confinement. This analysis shows the thermal plasma energy confinement enhancement factor to be independent of the heating mix, for the range of conditions explored. Moreover, the response of the global confinement to changes in density and power were also independent of heating mix, and were reflecting the changes in the pedestal, which is consistent with the profiles being globally stiff. Consistently, the pedestal characteristics (pressure and width) and their dependences were the same with NB or with predominant IC heating.