Publications (14)27.57 Total impact
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Article: Electroacupuncture inhibition of hyperalgesia in an inflammatory pain rat model: involvement of distinct spinal serotonin and norepinephrine receptor subtypes.
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ABSTRACT: Although acupuncture analgesia is well documented, its mechanisms have not been thoroughly clarified. We previously showed that electroacupuncture (EA) activates supraspinal serotonin- and norepinephrine-containing neurones that project to the spinal cord. This study investigates the involvement of spinal alpha(2)-adrenoceptors (α2-ARs) and 5-hydroxytryptamine (serotonin) receptors (5-HTRs) in EA effects on an inflammatory pain rat model. Inflammatory hyperalgesia was induced by injecting complete Freund's adjuvant (CFA, 0.08 ml) into the plantar surface of one hind paw and assessed by paw withdrawal latency (PWL) to a noxious thermal stimulus. The selective α2a-AR antagonist BRL-44408, α2b-AR antagonist imiloxan hydrochloride, 5-HT2B receptor (5-HT2BR) antagonist SB204741, 5-HT3R antagonist LY278584, or 5-HT1AR antagonists NAN-190 hydrobromide, or WAY-100635 were intrathecally administered 20 min before EA or sham EA, which was given 2 h post-CFA at acupoint GB30. EA significantly increased PWL compared with sham [7.20 (0.46) vs 5.20 (0.43) s]. Pretreatment with α2a-AR [5.35 (0.45) s] or 5-HT1AR [5.22 (0.38) s] antagonists blocked EA-produced anti-hyperalgesia; α2b-AR, 5-HT2BR, and 5-HT3R antagonist pretreatment did not. Sham plus these antagonists did not significantly change PWL compared with sham plus vehicle, indicating that the antagonists had little effect on PWL. Immunohistochemical staining demonstrated that α2a-ARs are on primary afferents and 5-HT1ARs are localized in N-methyl-d-aspartic acid (NMDA) subunit NR1-containing neurones in the spinal dorsal horn. The data show that α2a-ARs and 5-HT1ARs are involved in the EA inhibition of inflammatory pain and that the NMDA receptors are involved in EA action.BJA British Journal of Anaesthesia 05/2012; 109(2):245-52. · 4.24 Impact Factor -
Article: Electroacupuncture alleviates affective pain in an inflammatory pain rat model.
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ABSTRACT: Pain has both sensory-discriminative and emotional-affective dimensions. Previous studies demonstrate that electroacupuncture (EA) alleviates the sensory dimension but do not address the affective. An inflammatory pain rat model, produced by a complete Freund adjuvant (CFA) injection into the hind paw, was combined with a conditioned place avoidance (CPA) test to determine whether EA inhibits spontaneous pain-induced affective response and, if so, to study the possibility that rostral anterior cingulate cortex (rACC) opioids underlie this effect. Male Sprague-Dawley rats (250-275 g, Harlan) were used. The rats showed place aversion (i.e. affective pain) by spending less time in a pain-paired compartment after conditioning than during a preconditioning test. Systemic non-analgesic morphine (0.5 and 1.0 mg/kg, i.p.) inhibited the affective reaction, suggesting that the affective dimension is underpinned by mechanisms different from those of the sensory dimension of pain. Morphine at 0.5 and at 1 mg/kg did not induce reward. Rats given EA treatment before pain-paired conditioning at GB 30 showed no aversion to the pain-paired compartment, indicating that EA inhibited the affective dimension. EA treatment did not produce reward or aversive effect. Intra-rACC administration of D-Phe-Cys-Tyr-D-Trp-Orn-Thr-Pen-Thr amide (CTOP), a selective mu opioid receptor antagonist, but not norbinaltorphimine (nor-BNI), a selective kappa opioid receptor antagonist, blocked EA inhibition of the affective dimension. These data demonstrate that EA activates opioid receptors in the rACC to inhibit pain-induced affective responses and that EA may be an effective therapy for both the sensory-discriminative and the affective dimensions of pain.European journal of pain (London, England) 02/2012; 16(2):170-81. · 3.37 Impact Factor -
Article: Overview of physics results from NSTX
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ABSTRACT: In the last two experimental campaigns, the low aspect ratio NSTX has explored physics issues critical to both toroidal confinement physics and ITER. Experiments have made extensive use of lithium coatings for wall conditioning, correction of non-axisymmetric field errors and control of n = 1 resistive wall modes (RWMs) to produce high-performance neutral-beam heated discharges extending to 1.7 s in duration with non-inductive current fractions up to 0.7. The RWM control coils have been used to trigger repetitive ELMs with high reliability, and they have also contributed to an improved understanding of both neoclassical tearing mode and RWM stabilization physics, including the interplay between rotation and kinetic effects on stability. High harmonic fast wave (HHFW) heating has produced plasmas with central electron temperatures exceeding 6 keV. The HHFW heating was used to show that there was a 20–40% higher power threshold for the L–H transition for helium than for deuterium plasmas. A new diagnostic showed a depletion of the fast-ion density profile over a broad spatial region as a result of toroidicity-induced Alfvén eigenmodes (TAEs) and energetic-particle modes (EPMs) bursts. In addition, it was observed that other modes (e.g. global Alfvén eigenmodes) can trigger TAE and EPM bursts, suggesting that fast ions are redistributed by high-frequency AEs. The momentum pinch velocity determined by a perturbative technique decreased as the collisionality was reduced, although the pinch to diffusion ratio, Vpinch/χ, remained approximately constant. The mechanisms of deuterium retention by graphite and lithium-coated graphite plasma-facing components have been investigated. To reduce divertor heat flux, a novel divertor configuration, the 'snowflake' divertor, was tested in NSTX and many beneficial aspects were found. A reduction in the required central solenoid flux has been realized in NSTX when discharges initiated by coaxial helicity injection were ramped in current using induction. The resulting plasmas have characteristics needed to meet the objectives of the non-inductive start-up and ramp-up program of NSTX.Nuclear Fusion 08/2011; 51(9):094011. · 4.09 Impact Factor -
Article: Overview of results from the National Spherical Torus Experiment (NSTX)
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ABSTRACT: The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high b operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance li [?] 0.4 with strong shaping (k [?] 2.7, d [?] 0.8) with bN approaching the with-wall b-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction fNI [?] 71%. Instabilities driven by super-Alfvenic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvenic. Linear toroidal Alfven eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with b above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.Nuclear Fusion. 01/2009; 49(10):104016. -
Article: Development of completely solenoidless tokamak operation in JT-60U
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ABSTRACT: Plasma current start-up to 100 kA was achieved successfully in the JT-60U tokamak without the use of the centre solenoid (completely solenoidless tokamak operation). Only poloidal field coils located on the outboard side of the torus were used, in combination with strong ionization by electron cyclotron (EC) power. The presence of a field null was not necessary for plasma current start-up, but the flux conversion efficiency was low in such a case. In a nearly solenoidless start-up, low neutral pressures were favoured, and the optimum location of the EC resonance was slightly to the high field side of the vacuum vessel centre. The required EC power for efficient utilization of flux swing in JT-60U was about 1 MW. A plasma current of 260 kA was maintained for 1 s by NB only, and plasma current ramp-up from 215 to 310 kA was achieved by EC and neutral beam (NB) only (without lower hybrid current drive (LHCD)). However, the ramp-up efficiency was much lower compared with LHCD. Recharging of the centre solenoid was observed with only counter and perpendicular NB injection, indicating bootstrap overdrive. Integration of these elements can lead to the achievement of a completely solenoidless tokamak operation.Nuclear Fusion 01/2006; 46(2):207. · 4.09 Impact Factor -
Article: Results of NSTX heating experiments
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ABSTRACT: The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved β<sub>T</sub> of 32%. A variety of MHD phenomena have been observed to limit β. NSTX has now begun addressing τ<sub>E</sub> scaling, β limits, and current drive issues. During the NBI heating experiments, a broad T<sub>i</sub> profile with T<sub>i</sub> up to 2 keV, T<sub>i</sub>>T<sub>e</sub> and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, τ<sub>E</sub> is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of τ<sub>E</sub> after the transition in some conditions. During HHFW heating, T<sub>e</sub>>T<sub>i</sub> and T<sub>e</sub> up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I<sub>p</sub>, high q, and high β<sub>p</sub>.IEEE Transactions on Plasma Science 03/2003; · 1.17 Impact Factor -
Article: Overview of JET results
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ABSTRACT: Scientific and technical activities on JET focus on the issues likely to affect the ITER design and operation. Our understanding of the ITER reference mode of operation, the ELMy H-mode, has progressed significantly. The extrapolation of ELM size to ITER has been re-evaluated. Neoclassical tearing modes have been shown to be meta-stable in JET, and their beta limits can be raised by destabilization (modification) of sawteeth by ion cyclotron radio frequency heating (ICRH). Alpha simulation experiments with ICRH accelerated injected 4 (He) beam ions provide a new tool for fast particle and magnetohydrodynamic studies, with up to 80-90% of plasma heating by fast 4 He ions. With or without impurity seeding, a quasi-steady-state high confinement (H-98 = 1), high density(n(e)/n(GW) = 0.9-1) and high beta (betaN = 2) ELMy H-mode has been achieved by operating near the ITER triangularity ( similar to 0.40-0.5) and safety factor (q(95) similar to 3), at Z(eff) similar to 1.5-2. In advanced tokamak (AT) scenarios, internal transport barriers (ITBs) are now characterized in real time with a new criterion, rhoT(*). Tailoring of the current profile with T lower hybrid current drive provides reliable access to a variety of q profiles, lowering access power for barrier formation. Rational q surfaces appear to be associated with ITB formation. Alfven cascades were observed in reversed shear plasmas, providing identification of q profile evolution. Plasmas with 'current holes' were observed and modelled. Transient high confinement AT regimes with H-89 = 3.3, beta(N) = 2.4 and ITER-relevant q < 5 were achieved with reversed magnetic shear. Quasi-stationary ITBs are developed with full non-inductive current drive, including similar to 50% bootstrap current. A record duration of ITBs was achieved, up to 11 s, approaching the resistive time. For the first time, pressure and current profiles of AT regimes are controlled by a real-time feedback system, in separate experiments. Erosion and co-deposition studies with a quartz micro-balance show reduced co-deposition. Measured divertor thermal loads during disruptions in JET could modify ITER assumptions.Nuclear Fusion 01/2003; · 4.09 Impact Factor -
Article: Impact of MHD equilibrium input variations on the high beta stability boundaries of NSTX
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ABSTRACT: Ideal MHD stability limits of anticipated plasma configurations for NSTX [Ono M. et al 2000 Nucl. Fusion 40 557] and the dependence on the parameters defining the MHD equilibrium are evaluated. The study provides a quantitative computational evaluation of the stability limit variations induced by changes to the equilibrium of NSTX high β plasmas. The analysis is based on a reference free-boundary equilibrium with β = 41.5%, monotonic safety factor q profile (qa = 12.1, q0 = 2.8) and broad pressure profile p (peaking factor Fp≡p(0)/p = 1.7). On this reference target local variation of the plasma boundary, and the safety factor q and pressure profiles p are imposed. Localized inflection of the outboard plasma boundary, produced by near field effects from poloidal shaping field coils, weaken the stability due to the destabilization of high n ballooning modes. Variation of the q profile at different radial locations can also degrade stability. Both experimental profiles from existing tokamaks and spherical torus machines and profiles generated from transport modelling of anticipated neutral beam heated plasmas are used. Degraded stability is found at increasing pressure peaking factor due to the destabilization of n = 1 kink/ballooning modes. Direct access to the second region of stability is found in certain configurations and, for the entire set of variations considered, the lower calculated β limit values are still in the range of 20.0% without considering the stabilizing effect of the passive conducting structures.Nuclear Fusion 04/2002; 42(4):418. · 4.09 Impact Factor -
Article: Non-inductive current generation in NSTX using coaxial helicity injection
04/2001; -
Conference Proceeding: High performance plasmas on the National Spherical Torus Experiment
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ABSTRACT: The National Spherical Torus Experiment has produced toroidal plasmas at low aspect ratio (A = R/a = 0.86 m/0.68 m ∼ 1.3, where R is the major radius and a is the minor radius of the torus) with plasma currents of 1.4 MA. The rapid development of the machine has led to very exciting physics results during the first full year of physics operation. Pulse lengths in excess of 0.5 s have been obtained with inductive current drive. Up to 4 MW of high harmonic fast wave (HHFW) heating power has been applied with 6 MW planned. Using only 2 MW of HHFW heating power clear evidence of electron heating is seen with HHFW, as observed by the multi point Thomson scattering diagnostic. A noninductive current drive concept known as coaxial helicity injection (CHI) has driven 260 kA of toroidal current. A neutral beam heating power of 5 MW has been injected. Plasmas with β<sub>1</sub> (= 2μ<sub>0</sub> < p > /B<sup>2</sup> = a measure of magnetic confinement efficiency) of 22% have been achieved, as calculated using the EFIT equilibrium reconstruction code. β limiting phenomena have been observed, and the maximum β<sub>1</sub> scales with I<sub>p</sub>/B<sub>1.</sub> High frequency (> MHz) magnetic fluctuations have been observed. H-mode plasmas are observed with confinement times of > 100 s. Beam heated plasmas show energy confinement times in excess of those predicted by empirical scaling expressions. Ion temperatures in excess of 2.0 keV have been measured, and the power balance suggests that the power loss from the ions to the electrons may exceed the calculated classical input power to the ions.Pulsed Power Plasma Science, 2001. PPPS-2001. Digest of Technical Papers; 02/2001 -
Article: Disappearance of giant ELMs and appearance of minute grassy ELMs in JT-60U high-triangularity discharges
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ABSTRACT: In JT-60U H-mode plasmas, giant (type I) ELMs disappear and minute grassy ELMs appear when triangularity δ, edge safety factor q95 and βp are high enough. Complete suppression of giant ELMs was observed at δ0.45, q956 and βp1.6. At higher δ (0.54), giant ELMs can disappear at a lower q95 (~4.0). In the grassy ELMy H-mode, edge temperature and pressure can be higher than those in giant ELMy H-mode and a favourable confinement can be sustained without an increase of the impurity concentration. An edge stability analysis suggests that the edge plasma is accessing the second stability regime of the high n ballooning mode in the grassy ELMy discharges.Plasma Physics and Controlled Fusion 05/2000; 42(5A):A247. · 2.42 Impact Factor -
Article: High performance plasmas on the National Spherical Ttorus Experiment
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ABSTRACT: The National Spherical Torus Experiment has produced toroidal plasmas at low aspect ratio (A = R/a = 0.86m/0.68m similar to 1.3, where R is the major radius and a is the minor radius of the torus) with plasma currents of 1.4MA. The rapid development of the machine has led to very exciting physics results during the first full year of physics operation. Pulse lengths in excess of 0.5s have been obtained with inductive current drive. Up to 4MW of High Harmonic Fast Wave (HHFW) heating power has been applied with 6MW planned. Using only 2MW of HHFW heating power clear evidence of electron heating is sden with HHFW, as observed by the multi point Thomson scattering diagnostic. A non-inductive current drive concept known as Coaxial Helicity Injection (CHI) has driven 260kA of toroidal current. Neutral beam heating power of 5MW has been injected. Plasmas with beta(t) (=2mu(0) /B-2 = a measure of magnetic confinement efficiency) of 22% have been achieved, as calculated using the EFIT equilibrium reconstruction code. beta limiting phenomena have been observed, and the maximum beta, scales with I-p/aB(t). High frequency (>MHz) magnetic fluctuations have been observed. H-mode plasmas are observed with confinement times of > 100ms. Beam heated plasmas show energy confinement times in excess of those predicted by empirical scaling expressions. Ion temperatures in excess of 2.0keV have been measured, and power balance suggests that the power loss from the ions to the electrons may exceed the calculated classical input power to the ions.Ppps-2001: Pulsed Power Plasma Science 2001, Vols I and Ii, Digest of Technical Papers. -
Article: Initial physics results from the National Spherical Torus Experiment
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ABSTRACT: The mission of the National Spherical Torus Experiment (NSTX) is to extend the understanding of toroidal physics to low aspect ratio (R/a similar or equal to1.25) in low collisionality regimes. NSTX is designed to operate with up to 6 MW of high harmonic fast wave (HHFW) heating and current drive, 5 MW of neutral beam injection (NBI) and co-axial helicity injection (CHI) for noninductive startup. Initial experiments focused on establishing conditions that will allow NSTX to achieve its aims of simultaneous high beta (t) and high-bootstrap current fraction, and to develop methods for noninductive operation, which will be necessary for Spherical Torus power plants. Ohmic discharges with plasma currents up to 1 MA and with a range of shapes and configurations were produced. Density limits in deuterium and helium reached 80% and 120% of the Greenwald limit, respectively. Significant electron heating was observed with up to 2.3 MW of HHFW. Up to 270 kA of toroidal current for up to 200 ms was produced noninductively using CHI. Initial NBI experiments were carried out with up to two beam sources (3.2 MW). Plasmas with stored energies of up to 140 kJ and beta (t)=21% were produced. (C) 2001 American Institute of Physics.Physics of Plasmas. 8(5):1977-1987. -
Article: Dependence of H-mode pedestal parameters on plasma magnetic geometry
Plasma Physics and Controlled Fusion, v.44, A273-A278 (2002).
Top co-authors
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Institutions
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2012
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University of Maryland, Baltimore
- Department of Medicine
Baltimore, MD, USA
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2000–2011
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General Atomics
San Diego, CA, USA
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