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Oscillating Vertical Magnetic Dipole Above a Conducting Half-Space

Description: The electromagnetic field produced by a vertical oscillating magnetic dipole above a plane conducting earth is obtained in integral form. An exact solution in closed form is obtained for the case in which the dipole and the point of observation are both located on the surface of the earth. (auth)
Date: April 1, 1961
Creator: Wesley, J. P.
Partner: UNT Libraries Government Documents Department

1. 5 megawatt dc chopper power supplies for plasma shape control on Doublet III

Description: The Doublet III device is designed to study noncircular plasmas, including doublet and dee-shaped cross-sections. The plasma shape is determined by a system of 24 field-shaping coils which surround the vacuum vessel. Control of the magnetic flux linking these coils allows the plasma shape to be varied and controlled. This paper describes the high-speed dc chopper which is a major component of the field-shaping coil power system. The high-speed dc choppers, with a frequency response of up to 5 kHz and a switching power capability of 1.5 megawatts are used for fine tuning and feedback control of the plasma position and shape. The design and operation of two 1.5 megawatt, 3 kHz choppers used on closed loop plasma control experiments will be presented.
Date: November 1, 1979
Creator: Rock, P.J. & Wesley, J.C.
Partner: UNT Libraries Government Documents Department

Shaping and characteristics of ohmically heated noncircular plasmas in Doublet III

Description: Ohmically heated dee, droplet and doublet plasmas with vertical elongations of up to 3.2 have been produced in Doublet III. Doublet configurations are now obtained by controlled merging of the two current channels of a droplet discharge. All discharges have low levels of impurities and central radiated power and achieve high density relative to B/sub T//R scaling. The energy confinement follows the standard circular cross-section Alcator scaling law. The lack of significant improvement in the confinement due to vertical elongation is consistent with the high degree of current peaking inferred from magnetic measurements.
Date: June 1, 1980
Creator: Wesley, J.C.; Angel, T. & Armentrout, C.J.
Partner: UNT Libraries Government Documents Department

Doublet III limiter performance and implications for mechanical design and material selection for future limiters

Description: The plasma limiter system for Doublet III is described. Initially, high-Z materials, Ta-10W for the primary limiter and Mo for the backup limiters, were selected as the most attractive metallic candidates from the standpoint of thermal and structural properties. For the purpose of evaluating the effect of material Z on plasma performance, the nonmagnetic, Ni-base alloy Inconel X-750 was selected for a medium-Z limiter material. Graphite, a low-Z material, will likely be the next limiter material for evaluation. Design and material selection criteria for the different Z ranges are presented. The performance of the high-Z limiters in Doublet III is reviewed for an operation period that included approximately 5000 plasma shots. Changes in surface appearance and metallurgical changes are characterized. Discussion is presented on how and to what extent the high-Z elements affected the performance of the plasma based on theory and measurements in Doublet III. The fabrication processes for the Inconel X-750 limiters are summarized, and, last, observations on early performance of the Inconel limiters are described. (MOW)
Date: October 1, 1979
Creator: Sabado, M.M.; Marcus, F.B.; Trester, P.W. & Wesley, J.C.
Partner: UNT Libraries Government Documents Department

X-irradiation Effects on the Action Potentials of Frog Sciatic Nerves Inhibited by Carbon Monoxide and Ouabain

Description: The response of frog sciatic nerve action potentials to x-irradiation and metabolic (carbon monoxide) or transport (ouabain) inhibition was determined in an attempt to further identify the nature of radiation insult to nervous tissue. Carbon monoxide, ouabain (2 X 10-5 M), and nitrogen anoxia were shown to produce a near linear decline in action potential amplitude. The carbon monoxide and nitrogen inhibitions of activity were reversible in air; the carbon monoxide inhibition was light reversible. Ouabain inhibition was partially reversible by soaking the nerve in aerated Ringer's. Application of 120 kv x-rays (75 Kr at 4.9 Kr/min) to nerves during the linear decline in spike amplitude brought about a marked enhancement (146%) of inhibition by 99% CO/l% 02, nitrogen (136%), and ouabain (265%). All bhanges were shown to be statistically significant by a regression analysis. However, x-irradiation did not appear to alter the air reversibility of carbon monoxide and nitrogen inhibitions nor the reversibility in Ringerts of the ouabain inhibition. Additionally x-irradiation completely blocked light reversal of 98% CO/2% 02 inhibition and produced a decline in activity. A possible interpretation of these results is a compensation for radiation action at this dosage requiring metabolism and ion pump activity.
Date: December 1971
Creator: Thompson, Wesley J.
Partner: UNT Libraries

The Evaluation of Task Preference on Reinforcer Efficacy

Description: Stimulus preference assessments have determined high and low preferred items that increase the rate of frequency of responding for various skills. Within applied settings, high preferred items may not attain the same reinforcing value across tasks which might decrease responding. The preference of the task might have an effect on reinforcer efficacy that is being presented. The purpose of the current study is to evaluate changes in reinforcer efficacy as a function of preference for the task. Three children diagnosed with ASD participated in the study. HP/LP items and HP/LP tasks were identified through paired-choice assessments, and each item was presented as a consequence for each task in a counterbalanced multi-element format. Results indicated that preference for the task had little effect of the rate of responding across items.
Date: December 2014
Creator: Lowery, Wesley J.
Partner: UNT Libraries

Negative ion photoelectron spectroscopy of P₂N₃⁻: electron affinity and electronic structures of P₂N₃˙

Description: This article reports a negative ion photoelectron spectroscopy (NIPES) and ab initio study of the recently synthesized planar aromatic inorganic ion P₂N₃⁻, to investigate the electronic structures of P₂N₃⁻ and its neutral P₂N₃˙ radical.
Date: April 5, 2016
Creator: Hou, Gao-Lei; Chen, Bo; Transue, Wesley J.; Hrovat, David A.; Cummins, Christopher C.; Borden, Weston T. et al.
Partner: UNT College of Arts and Sciences

Conceptual design summary for modifying Doublet III to a large dee-shaped configuration

Description: The Doublet III tokamak is to be reconfigured by replacing its indented (doublet) vacuum vessel with a larger one of a dee-shaped cross section. This change will permit significantly larger elongated plasmas than is presently possible and will allow higher plasma current (up to 5 MA) and anticipated longer confinement time. Reactor relevant values of stable beta and plasma pressure are predicted. This modification, while resulting in a significant change in capability, utilizes most of the existing coils, structure, systems and facility.
Date: May 1, 1983
Creator: Davis, L.G.; Gallix, R.; Luxon, J.L.; Mahdavi, M.A.; Puhn, F.A.; Rock, P.J. et al.
Partner: UNT Libraries Government Documents Department

ITER-EDA physics design requirements and plasma performance assessments

Description: Physics design guidelines, plasma performance estimates, and sensitivity of performance to changes in physics assumptions are presented for the ITER-EDA Interim Design. The overall ITER device parameters have been derived from the performance goals using physics guidelines based on the physics R&D results. The ITER-EDA design has a single-null divertor configuration (divertor at the bottom) with a nominal plasma current of 21 MA, magnetic field of 5.68 T, major and minor radius of 8.14 m and 2.8 m, and a plasma elongation (at the 95% flux surface) of {approximately}1.6 that produces a nominal fusion power of {approximately}1.5 GW for an ignited burn pulse length of {ge}1000 s. The assessments have shown that ignition at 1.5 GW of fusion power can be sustained in ITER for 1000 s given present extrapolations of H-mode confinement ({tau}{sub E} = 0.85 {times} {tau}{sub ITER93H}), helium exhaust ({tau}*{sub He}/{tau}{sub E} = 10), representative plasma impurities (n{sub Be}/n{sub e} = 2%), and beta limit [{beta}{sub N} = {beta}(%)/(I/aB) {le} 2.5]. The provision of 100 MW of auxiliary power, necessary to access to H-mode during the approach to ignition, provides for the possibility of driven burn operations at Q = 15. This enables ITER to fulfill its mission of fusion power ({approximately} 1--1.5 GW) and fluence ({approximately}1 MWa/m{sup 2}) goals if confinement, impurity levels, or operational (density, beta) limits prove to be less favorable than present projections. The power threshold for H-L transition, confinement uncertainties, and operational limits (Greenwald density limit and beta limit) are potential performance limiting issues. Improvement of the helium exhaust ({tau}*{sub He}/{tau}{sub E} {le} 5) and potential operation in reverse-shear mode significantly improve ITER performance.
Date: July 1, 1996
Creator: Uckan, N.A.; Galambos, J.; Wesley, J.; Boucher, D.; Perkins, F.; Post, D. et al.
Partner: UNT Libraries Government Documents Department

Plasma equilibrium control in doublet III

Description: The control signals used for the flux surface position regulation are derived from measurements of the poloidal field and flux obtained with an array of sensors located immediately outside the vacuum vessel. The close proximity of the sensors to the plasma surface (less than or equal to 0.2a) allows the position of the plasma surface to be accurately computed with simple analog circuitry. The short time constant of the resistive vessel allows for stable high-gain closed loop control without elaborate pole-zero compensation. The small-signal response time is typically less than 10 msec. Control of the plasma position to +-0.5 cm is routine.
Date: October 1, 1980
Creator: Stambaugh, R.; Adcock, S.; Callis, R.; deGrassie, J.; Luxon, J.; Rock, P. et al.
Partner: UNT Libraries Government Documents Department

TSC plasma halo simulation of a DIII-D vertical displacement episode

Description: A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench, and current quench. Inclusion of a 1-to 4-eV halo surrounding the main plasma was found to be necessary to match simulation and experimental results for plasma current decay, trajectory, toroidal and poloidal vessel currents, and magnetic probe and flux loop values for the entire VDE.
Date: January 1, 1993
Creator: Sayer, R. O.; Peng, Y. K. M.; Jardin, S. C.; Kellman, A. G. & Wesley, J. C.
Partner: UNT Libraries Government Documents Department

ITER physics-safety interface: models and assessments

Description: Plasma operation conditions and physics requirements to be used as a basis for safety analysis studies are developed and physics results motivated by safety considerations are presented for the ITER design. Physics guidelines and specifications for enveloping plasma dynamic events for Category I (operational event), Category II (likely event), and Category III (unlikely event) are characterized. Safety related physics areas that are considered are: (i) effect of plasma on machined and safety (disruptions, runaway electrons, fast plasma shutdown) and (ii) plasma response to ex-vessel LOCA from first wall providing a potential passive plasma shutdown due to Be evaporation. Physics models and expressions developed are implemented in safety analysis code (SAFALY, couples 0-D dynamic plasma model to thermal response of the in-vessel components). Results from SAFALY are presented.
Date: October 1, 1996
Creator: Uckan, N.A.; Putvinski, S.; Wesley, J.; Bartels, H-W.; Honda, T.; Amano, T. et al.
Partner: UNT Libraries Government Documents Department

Studies of Impurity Assimilation During Massive Argon Gas Injection in DIII-D

Description: Fast shutdown of discharges using massive gas injection (MGI) is a promising technique for reducing tokamak wall damage during disruptions [1]. An outstanding concern, however, is the generation of runaway electrons (RE) during the shutdown. Although RE formation observed during MGI in present-day experiments is quite small (typically <1% of the main plasma current I{sub p} in DIII-D), it is thought that even this small RE current could be amplified to significant levels in reactor-scale tokamaks such as ITER [2]. It is expected that complete collisional suppression of any potential RE amplification during the CQ can be achieved for suppression parameters {gamma}{sub crit} {triple_bond} E{sub crit}/E{sub {psi}} > 1, where E{sub crit} = [2{pi}e{sup 3}ln{Lambda}(2n{sub e} + n{sub B})]/mc{sup 2} is the critical electric field [2] and E{sub {psi}} {approx} -[({mu}{sub 0}l{sub i})/4{pi}][-({partial_derivative}I{sub p}/{partial_derivative}t)+ {alpha}{sub L}(I{sub W}/{tau}{sub W})] is the toroidal electric field resulting from the decay of the plasma current I{sub p}. n{sub e} is the free electron density, n{sub B} is the bound electron density, {alpha}{sub L} {approx} 2[ln(8R/r{sub w})-2]/l{sub i} is the ratio of external (outside conducting wall) to internal (inside conducting wall) self-inductance, I{sub w} is the wall current, and {tau}{sub w} is the wall time. The densities required to achieve {gamma}{sub crit} > 1 are typically quite large, e.g. n{sub tot} {triple_bond} n{sub e} + n{sub B}/2 {approx} 10{sup 16} cm{sup -3} for DIII-D. To have a possibility of achieving the required density in the DIII-D plasma (with volume V{sub p} {approx} 20 m{sup 3}), an MGI system using argon must be able to deliver of order 10{sup 22} argon atoms to the plasma within the shutdown timescale of about 10 ms.
Date: June 27, 2007
Creator: Hollmann, E; Jernigan, T; Parks, P; Baylor, L; Boedo, J; Combs, S et al.
Partner: UNT Libraries Government Documents Department

DIII-D Studies of Massive Gas Injection Fast Shutdowns for Disruption Mitigation

Description: Injection of massive quantities of gas is a promising technique for fast shutdown of ITER for the purpose of avoiding divertor and first wall damage from disruptions. Previous experiments using massive gas injection (MGI) to terminate discharges in the DIII-D tokamak have demonstrated rapid shutdown with reduced wall heating and halo currents (relative to natural disruptions) and with very small runaway electron (RE) generation [1]. Figure 1 shows time traces which give an overview of shutdown time scales. Typically, of order 5 x 10{sup 22} Ar neutrals are fired over a pulse of 25 ms duration into stationary (non-disrupting) discharges. The observed results are consistent with the following scenario: within several ms of the jet trigger, sufficient Ar neutrals are delivered to the plasma to cause the edge temperature to collapse, initiating the inward propagation of a cold front. The exit flow of the jet [Fig. 1(a)] has a {approx} 9 ms rise time; so the quantity of neutrals which initiates the edge collapse is small (<10{sup 20}). When the cold front reaches q {approx} 2 surface, global magnetohydrodynamic (MHD) modes are destabilized [2], mixing hot core plasma with edge impurities. Here, q is the safety factor. Most (>90%) of the plasma thermal energy is lost via impurity radiation during this thermal quench (TQ) phase. Conducted heat loads to the wall are low because of the cold edge temperature. After the TQ, the plasma is very cold (of order several eV), so conducted wall (halo) currents are low, even if the current channel contacts the wall. The plasma current profile broadens and begins decaying resistively. The decaying current generates a toroidal electric field which can accelerate REs; however, RE beam formation appears to be limited in MGI shutdowns. Presently, it is thought that the conducted heat flux and halo current ...
Date: June 19, 2006
Creator: Hollmann, E; Jernigan, T; Antar, G; Bakhtiari, M; Boedo, J; Combs, S et al.
Partner: UNT Libraries Government Documents Department

DIII-D Studies of Massive Gas Injection Fast Shutdowns for Disruption Mitigation

Description: Injection of massive quantities of gas is a promising technique for fast shutdown of ITER for the purpose of avoiding divertor and first wall damage from disruptions. Previous experiments using massive gas injection (MGI) to terminate discharges in the DIII-D tokamak have demonstrated rapid shutdown with reduced wall heating and halo currents (relative to natural disruptions) and with very small runaway electron (RE) generation [1]. Figure 1 shows time traces which give an overview of shutdown time scales. Typically, of order 5 x 10{sup 22} Ar neutrals are fired over a pulse of 25 ms duration into stationary (non-disrupting) discharges. The observed results are consistent with the following scenario: within several ms of the jet trigger, sufficient Ar neutrals are delivered to the plasma to cause the edge temperature to collapse, initiating the inward propagation of a cold front. The exit flow of the jet [Fig. 1(a)] has a {approx} 9 ms rise time; so the quantity of neutrals which initiates the edge collapse is small (<10{sup 20}). When the cold front reaches q {approx} 2 surface, global magnetohydrodynamic (MHD) modes are destabilized [2], mixing hot core plasma with edge impurities. Here, q is the safety factor. Most (>90%) of the plasma thermal energy is lost via impurity radiation during this thermal quench (TQ) phase. Conducted heat loads to the wall are low because of the cold edge temperature. After the TQ, the plasma is very cold (of order several eV), so conducted wall (halo) currents are low, even if the current channel contacts the wall. The plasma current profile broadens and begins decaying resistively. The decaying current generates a toroidal electric field which can accelerate REs; however, RE beam formation appears to be limited in MGI shutdowns. Presently, it is thought that the conducted heat flux and halo current ...
Date: September 29, 2006
Creator: Hollmann, E; Jernigan, T; Antar, G; Bakhtiari, M; Boedo, J; Combs, S et al.
Partner: UNT Libraries Government Documents Department

ITER plasma safety interface models and assessments

Description: Physics models and requirements to be used as a basis for safety analysis studies are developed and physics results motivated by safety considerations are presented for the ITER design. Physics specifications are provided for enveloping plasma dynamic events for Category I (operational event), Category II (likely event), and Category III (unlikely event). A safety analysis code SAFALY has been developed to investigate plasma anomaly events. The plasma response to ex-vessel component failure and machine response to plasma transients are considered.
Date: December 31, 1996
Creator: Uckan, N.A.; Bartels, H-W.; Honda, T.; Putvinski, S.; Amano, T.; Boucher, D. et al.
Partner: UNT Libraries Government Documents Department

ITER disruption modeling using TSC (Tokamak Simulation Code)

Description: Design of the ITER vacuum vessel (VV) is driven strongly by disruption-induced forces. We use the Tokamak Simulation Code (TSC) to model disruptions for the ITER physics phase (I{sub p} = 22 MA) and predict the time evolution of currents and forces on the VV. For a plasma vertically displaced to Z{sub axis} = {minus}1.0m before disruption and decaying at a rate of < dI{sub p}/dt > {approx equal} {minus}1.0MA/ms, the induced VV current peaks at 18 MA. The maximum radial VV force F{sub R} is 56 MN/rad; the maximum vertical force F{sub Z} is 5.4 MN/rad; and the maximum VV disruption pressure is 1.0 MPa. Variations in VV resistance (20 - 160 {mu}{Omega}) and < dI{sub p}/dt > (1 - 2.5 MA/ms) do not change F{sub R} significantly. The dependence of the forces on the initial plasma displacement and < dI{sub p}/dt > behavior, and the responses of other conducting structures are discussed. 2 refs., 6 figs.
Date: November 13, 1989
Creator: Sayer, R.O.; Peng, Y.K.M.; Wesley, J.C.; Jardin, S.C. (Oak Ridge National Lab., TN (USA); General Atomics, San Diego, CA (USA) & Princeton Univ., NJ (USA). Plasma Physics Lab.)
Partner: UNT Libraries Government Documents Department

Physics basis for the Fusion Ignition Research Experiment (FIRE)

Description: Understanding the properties of high gain (alpha-dominated) fusion plasmas in an advanced toroidal configuration is a critical issue that must be addressed to provide the scientific foundation for an attractive magnetic fusion reactor. The functional fusion plasma objectives for major next physics steps in magnetic fusion research can be described as: Burning Plasma Physics - The achievement and understanding of alpha-dominated plasmas that have characteristics similar to those expected in a fusion energy source, and Advanced Toroidal Physics - The achievement and understanding of bootstrap-current-dominated plasmas with externally controlled profiles and other characteristics (e.g. confinement and beta) similar to those expected in an attractive fusion system.
Date: July 7, 2000
Creator: Meade, D. M.; Thome, R. J.; Sauthoff, N. R.; Heitzenroeder, P. J.; Nelson, B. E.; Ulrickson, M.A et al.
Partner: UNT Libraries Government Documents Department

Physics Regimes in the Fusion Ignition Research Experiment (FIRE)

Description: Burning plasma science is recognized widely as the next frontier in fusion research. The Fusion Ignition Research Experiment (FIRE) is a design study of a next-step burning plasma experiment with the goal of developing a concept for an experimental facility to explore and understand the strong nonlinear coupling among confinement, magnetohydrodynamic (MHD) self-heating, stability, edge physics, and wave-particle interactions that is fundamental to fusion plasma behavior. This will require plasmas dominated by alpha heating (Q greater than or equal to 5) that are sustained for a duration comparable to characteristic plasma timescales (greater than or equal to 10) tau(subscript ''E''), approximately 4 tau(subscript ''He''), approximately 2 tau(subscript ''skin''). The work reported here has been undertaken with the objective of finding the minimum size (cost) device to achieve these physics goals.
Date: June 19, 2001
Creator: Meade, D.M.; S.C.Jardin; Kessel, C.E.; Ulrickson, M.A.; Schultz, J.H.; Rutherford, P.H. et al.
Partner: UNT Libraries Government Documents Department

TSC plasma halo simulation of a DIII-D vertical displacement episode

Description: A benchmark of the Tokamak Simulation Code (TSC) plasma halo model has been achieved by calibration against a DIII-D vertical displacement episode (VDE) consisting of vertical drift, thermal quench, and current quench. Inclusion of a 1-to 4-eV halo surrounding the main plasma was found to be necessary to match simulation and experimental results for plasma current decay, trajectory, toroidal and poloidal vessel currents, and magnetic probe and flux loop values for the entire VDE.
Date: January 1, 1993
Creator: Sayer, R.O.; Peng, Y.K.M. (Oak Ridge National Lab., TN (United States)); Jardin, S.C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kellman, A.G. & Wesley, J.C. (General Atomics, San Diego, CA (United States))
Partner: UNT Libraries Government Documents Department

Development of ITER 15 MA ELMy H-mode Inductive Scenario

Description: The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (&lt;0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation.
Date: October 16, 2008
Creator: Kessel, C. E.; Campbell, D.; Gribov, Y.; Saibene, G.; Ambrosino, G.; Casper, T. et al.
Partner: UNT Libraries Government Documents Department