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Lower hybrid wave electron heating experiments in Doublet IIA

Description: Experiments designed to heat electrons by Landau damping of waves at approximately twice the lower hybrid frequency have been carried out on Doublet IIA. This objective is in contrast to other lower hybrid experiments which are designed to heat ions using frequencies corresponding to the lower hybrid resonance frequency. Up to 500 kW of rf power was applied to discharge with approximately 100 kW ohmic input using parallel wavelengths chosen to optimize the spatial distribution of the power deposition based on linear or quasi-linear Landau damping.
Date: July 1, 1980
Creator: Freeman, R.L.; Luxon, J.L. & Chan, V.S.
Partner: UNT Libraries Government Documents Department

Effects of plasma shape and profiles on edge stability in DIII-D

Description: The results of recent experimental and theoretical studies concerning the effects of plasma shape and current and pressure profiles on edge instabilities in DIII-D are presented. Magnetic oscillations with toroidal mode number n {approx} 2--9 and a fast growth time {gamma}{sup {minus}1} = 20--150 {micro}s are often observed prior to the first giant type 1 ELM in discharges with moderate squareness. High n ideal ballooning second stability access encourages edge instabilities by facilitating the buildup of the edge pressure gradient and bootstrap current density which destabilize the intermediate to low n modes. Analysis suggests that discharges with large edge pressure gradient and bootstrap current density are more unstable to n > 1 modes. Calculations and experimental results show that ELM amplitude and frequency can be varied by controlling access to the second ballooning stability regime at the edge through variation of the squareness of the discharge shape. A new method is proposed to control edge instabilities by reducing access to the second ballooning stability regime at the edge using high order local perturbation of the plasma shape in the outboard bad curvature region.
Date: December 1, 1998
Creator: Lao, L.L.; Chan, V.S. & Chen, L.
Partner: UNT Libraries Government Documents Department

Development path of low aspect ratio tokamak power plants

Description: Recent advances in tokamak physics indicate the spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants. Full calculations of stability to kink and ballooning modes show the possibility of greater than 50% beta toroidal with the normalized beta as high as 10 and fully aligned 100% bootstrap current. Such beta values coupled with 2--3 T toroidal fields imply a pilot plant about the size of the present DIII-D tokamak could produce {approximately} 800 MW thermal, 160 MW net electric, and would have a ratio of gross electric power over recirculating power (Q{sub PLANT}) of 1.9. The high beta values in the ST mean that E x B shear stabilization of turbulence should be 10 times more effective in the ST than in present tokamaks, implying that the required high quality of confinement needed to support such high beta values will be obtained. The anticipated beta values are so high that the allowable neutron flux at the blanket sets the device size, not the physics constraints. The ST has a favorable size scaling so that at 2--3 times the pilot plant size the Q{sub PLANT} rises to 4--5, an economic range and 4 GW thermal power plants result. Current drive power requirements for 10% of the plasma current are consistent with the plant efficiencies quoted. The unshielded copper centerpost should have an adequate lifetime against nuclear transmutation induced resistance change and the low voltage, high current power supplies needed for the 12 turn TF coil appear reasonable. The favorable size scaling of the ST and the high beta mean that in large sizes, if the copper TF coil is replaced with a superconducting TF coil and a shield, the advanced fuel D-He{sup 3} could ...
Date: March 1, 1997
Creator: Stambaugh, R.D.; Chan, V.S. & Miller, R.L.
Partner: UNT Libraries Government Documents Department

Confinement and stability of DIII-D negative central shear discharges

Description: Negative central magnetic shear (NCS) discharges with {Beta}{sub N} {le} 4, H {le} 3, and up to 80% of the current non-inductively driven are reproducibly produced in the DIII-D tokamak. Strong peaking of T{sub i}, plasma rotation, and in some cases n{sub e} are observed inside the NCS region. Transport analysis shows that the core ion thermal diffusivity is substantially reduced and near the neoclassical value after the formation of the internal transport barrier. The negative central shear is necessary but not sufficient for the formation of this transport barrier. The power required for the formation appears to increase with the toroidal magnetic field. The high performance phase of H-mode NCS discharges often ends with an ELM-like collapse initiated from the edge whereas the L-mode discharges which have a more peaked pressure profile tend to end with a more global n = 1 MHD event.
Date: December 1, 1995
Creator: Lao, L.L.; Burrell, K.H. & Chan, V.S.
Partner: UNT Libraries Government Documents Department

Advances and Challenges in Computational Plasma Science

Description: Scientific simulation, which provides a natural bridge between theory and experiment, is an essential tool for understanding complex plasma behavior. Recent advances in simulations of magnetically-confined plasmas are reviewed in this paper with illustrative examples chosen from associated research areas such as microturbulence, magnetohydrodynamics, and other topics. Progress has been stimulated in particular by the exponential growth of computer speed along with significant improvements in computer technology.
Date: January 3, 2005
Creator: Tang, W.M. & Chan, V.S.
Partner: UNT Libraries Government Documents Department


Description: Green's-function techniques are used to calculate electron cyclotron current drive (ECCD) efficiency in general tokamak geometry in the low-collisionality regime. Fully relativistic electron dynamics is employed in the theoretical formulation. The high-velocity collision model is used to model Coulomb collisions and a simplified quasi-linear rf diffusion operator describes wave-particle interactions. The approximate analytic solutions which are benchmarked with a widely used ECCD model, facilitate time-dependent simulations of tokamak operational scenarios using the non-inductive current drive of electron cyclotron waves.
Date: March 1, 2003
Partner: UNT Libraries Government Documents Department

Advanced Fusion Power Plant Studies. Annual Report for 1999

Description: Significant progress in physics understanding of the reversed shear advanced tokamak regime has been made since the last ARIES-RS study was completed in 1996. The 1999 study aimed at updating the physics design of ARIES-RS, which has been renamed ARIES-AT, using the improved understanding achieved in the last few years. The new study focused on: Improvement of beta-limit stability calculations to include important non-ideal effects such as resistive wall modes and neo-classical tearing modes; Use of physics based transport model for internal transport barrier (ITB) formation and sustainment; Comparison of current drive and rotational flow drive using fast wave, electron cyclotron wave and neutral particle beam; Improvement in heat and particle control; Integrated modeling of the optimized scenario with self-consistent current and transport profiles to study the robustness of the bootstrap alignment, ITB sustainment, and stable path to high beta and high bootstrap fraction operation.
Date: January 1, 2000
Creator: Chan, V.S.; Chu, M.S.; Greenfield, C.M.; Kinsey, J.E. & al., et
Partner: UNT Libraries Government Documents Department

Current drive and profile control in low aspect ratio tokamaks

Description: The key to the theoretically predicted high performance of a low aspect ratio tokamak (LAT) is its ability to operate at very large plasma current*I{sub p}. The plasma current at low aspect ratios follows the approximate formula: I{sub p} {approximately} (5a{sup 2}B{sub t}/Rq{psi}) [(1 + {kappa}{sup 2})/2] [A/(A {minus} 1)] where A {quadruple_bond} R/a which was derived from equilibrium studies. For constant q{psi} and B{sub t}, I{sub p} can increase by an order of magnitude over the case of tokamaks with A {approx_gt} 2.5. The large current results in a significantly enhanced {beta}{sub t} ({quadruple_bond} {beta}{sub N}I{sub p}/aB{sub t}) possibly of order unity. It also compensates for the reduction in A to maintain the same confinement performance assuming the confinement time {tau} follows the generic form {approximately} HI{sub p}P{sup {minus}1}/{sup 2}R{sup 3}/{sup 2}{kappa}{sup 1}/{sup 2}. The initiation and maintenance of such a large current is therefore a key issue for LATs.
Date: July 1, 1995
Creator: Chan, V.S.; Chiu, S.C.; Lin-Liu, Y.R.; Miller, R.L. & Turnbull, A.D.
Partner: UNT Libraries Government Documents Department


Description: OAK B202 A CLASS OF HIGH EBP EQUILIBRIA IN STRONGLY FINITE ASPECT RATIO TOKAMAK PLASMAS. A class of very high poloidal beta ({beta}{sub p}) equilibria is exhibited in which {var_epsilon}{beta}{sub p} ({var_epsilon} is the inverse aspect ratio) exceeds analytic equilibrium limits previously anticipated from simplified large aspect ratio models, as well as previous experimental equilibrium limits. The extension in these limits is shown to be due to a combination of finite aspect ratio effects and strong cross section shaping. In a conventional size configuration, equilibria with {var_epsilon}{beta}{sub p} = 3 exist, which is 50% higher than both the previously observed record and the analytically predicted limit. For very low aspect ratio with {var_epsilon}{sup -1} = 1.4, typical of spherical tori, values of {var_epsilon}{beta}{sub p} = 5 are possible.
Date: August 1, 2002
Partner: UNT Libraries Government Documents Department

Applications of fast wave in spherical tokamaks

Description: In spherical tokamaks (ST), the magnetic field strength varies over a wide range across the plasma, and at high betas it deviates significantly from the 1/R dependence of conventional tokamaks. This, together with the high density expected in ST, poses challenging problems for RF heating and current drive. In this paper, the authors investigate the various possible applications of fast waves (FW) in ST. The adjoint technique of calculating current drive is implemented in the raytracing code CURRAY. The applicability of high harmonic and subharmonic FW to steady state ST is considered. They find that high harmonic FW tends to be totally absorbed before reaching the core and may be considered a candidate for off axis current drive while the subharmonic FW tends to be absorbed mainly in the core region and may be considered for central current drive. A difficult problem is the maintenance of current at the startup stage. In the bootstrap ramp-up scenario, the current ramp-up is mainly provided by the bootstrap current. Under this condition, the role of rf becomes mainly the sustainment of plasma through electron heating. Using a slab full-wave code SEMAL, the authors find that the ion-ion-hybrid mode conversion scheme is a promising candidate. The effect of possible existence of edge Alfven resonance and high harmonic cyclotron resonance is investigated and regimes of minimization of edge heating identified.
Date: April 1, 1997
Creator: Chiu, S.C.; Chan, V.S.; Lin-Liu, Y.R.; Miller, R.L.; Prater, R. & Politzer, P.
Partner: UNT Libraries Government Documents Department

Stable bootstrap-current driven equilibria for low aspect ratio tokamaks

Description: Low aspect ratio tokamaks can potentially provide a high ratio of plasma pressure to magnetic pressure {beta} and high plasma current I at a modest size, ultimately leading to a high power density compact fusion power plant. For the concept to be economically feasible, bootstrap current must be a major component of the plasma current. A high value of the Troyon factor {beta}{sub N} and strong shaping are required to allow simultaneous operation at high {beta} and high bootstrap current fraction. Ideal magnetohydrodynamic stability of a range of equilibria at aspect ratio 1.4 is systematically explored by varying the pressure profile and shape. The pressure and current profiles are constrained in such a way as to assure complete bootstrap current alignment. Both {beta}{sub N} and {beta} are defined in terms of the vacuum toroidal field. Equilibria with {beta}{sub N} {ge} 8 and {beta} - 35% to 55% exist which are stable to n = {infinity} ballooning modes, and stable to n = 0, 1,2,3 kink modes with a conducting wall. The dependence of {beta} and {beta}{sub N} with respect to aspect ratio is also considered.
Date: August 1, 1996
Creator: Miller, R.L.; Lin-Liu, Y.R.; Turnbull, A.D.; Chan, V.S.; Pearlstein, L.D.; Sauter, O. et al.
Partner: UNT Libraries Government Documents Department

DIII-D Advanced Tokamak Research Overview

Description: This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously {beta}{sub N}H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues.
Date: December 1, 1999
Creator: Chan, V.S.; Greenfield, C.M.; Lao, L.L.; Luce, T.C.; Petty, C.C. & Staebler, G.M.
Partner: UNT Libraries Government Documents Department

Modeling of Trapped Electron Effects on Electron Cyclotron Current Drive for Recent DIII-D Experiments

Description: Owing to its potential capability of generating localized non-inductive current, especially off-axis, Electron Cyclotron Current Drive (ECCD) is considered a leading candidate for current profile control in achieving Advanced Tokamak (AT) operation. In recent DIII-D proof-of-principle experiments [1], localized off-axis ECCD has been clearly demonstrated for first time. The measured current drive efficiency near the magnetic axis agrees well with predictions of the bounce-averaged Fokker-Planck theory [2,3]. However, the off-axis current drive efficiency was observed to exceed the theoretical results, which predict significant degradation of the current drive efficiency due to trapped electron effects. The theoretical calculations have been based on an assumption that the effective collision frequency is much smaller than the bounce frequency such that the trapped electrons are allowed to complete the banana orbit at all energies. The assumption might be justified in reactor-grade tokamak plasmas, in which the electron temperature is sufficiently high or the velocity of resonant electrons is much larger than the thermal velocity, so that the influence of collisionality on current drive efficiency can be neglected. For off-axis deposition in the present-day experiments, the effect of high density and low temperature is to reduce the current drive efficiency, but the increasing collisionality reduces the trapping of current-carrying electrons, leading to compensating increases in the current drive efficiency. In this work, we use the adjoint function formulation [4] to examine collisionality effects on the current drive efficiency.
Date: August 1999
Creator: Lin-Liu, Y. R.; Sauter, O.; Harvey, R. W.; Chan, V. S.; Luce, T. C. & Prater, R.
Partner: UNT Libraries Government Documents Department

Optimization and control of the plasma shape and current profile in noncircular cross-section tokamaks

Description: High-..beta.. equilibria which are stable to all ideal MHD modes are found by optimizing the plasma shape and current profile for doublets, up-down asymmetric dees, and symmetric dees. The ideal MHD stability of these equilibria for low toroidal mode number n is analyzed with a global MHD stability code, GATO. The stability to high-n modes is analyzed with a localized ballooning code, BLOON. The attainment of high ..beta.. is facilitated by an automated optimization search on shape and current parameters. The equilibria are calculated with a free-boundary equilibrium code using coils appropriate for the Doublet III experimental device. The optimal equilibria are characterized by broad current profiles with values of ..beta../sub poloidal/ approx. =1. Experimental realization of the shapes and current profiles giving the highest ..beta.. limits is explored with a 1 1/2-D transport code, which simulates the time evolution of the 2-D MHD equilibrium while calculating consistent current profiles from a 1-D transport model. Transport simulations indicate that nearly optimal shapes may be obtained provided that the currents in the field-shaping coils are appropriately programmed and the plasma current profile is sufficiently broad. Obtaining broad current profiles is possible by current ramping, neutral beam heating, and electron cyclotron heating. With combinations of these techniques it is possible to approach the optimum ..beta.. predicted by the MHD theory.
Date: June 1, 1980
Creator: Moore, R.W.; Bernard, L.C.; Chan, V.S.; Davidson, R.H.; Dobrott, D.R.; Helton, F.J. et al.
Partner: UNT Libraries Government Documents Department


Description: During 1998, the General Atomics (GA) ARIES-Spherical Torus (ST) team examined several critical issues related to the physics performance of the ARIES-ST design, and a number of suggestions were made concerning possible improvements in performance. These included specification of a reference plasma equilibrium, optimization about the reference equilibrium to achieve higher beta limits, examination of three possible schemes for plasma initiation, development of a detailed scenario for ramp-up of the plasma current and pressure to its full, final operating values, an assessment of the requirement for electron confinement, and several suggestions for divertor heat flux reduction. The reference equilibrium was generated using the TOQ code, with the specification of a 100%, self-consistent bootstrap current. The equilibrium has {beta} = 51%, 10% below the stability limit (a margin specified by the ARIES-ST study). In addition, a series of intermediate equilibria were defined, corresponding to the ramp-up scenario discussed. A study of the influence of shaping on ARIES-ST performance indicates that significant improvement in both kink and ballooning stability can be obtained by modest changes in the squareness of the plasma. In test equilibria the ballooning beta limit is increased from 58% to 67%. Also the maximum allowable plasma-wall separation for kink stability can be increased by 30%. Three schemes were examined for noninductive plasma initiation. These are helicity injection (HICD), electron cyclotron heating (ECH)-assisted startup, and inductive startup using only the external equilibrium coils. HICD startup experiments have been done on the HIT and CDX devices. ECH-assisted startup has been demonstrated on CDX-U and DIII-D. External coil initiation is based on calculations for a proposed DIII-D experiment. In all cases, plasma initiation and preparation of an approximately 0.3 MA plasma for ARIES-ST appears entirely feasible.
Date: April 1, 1999
Creator: CHAN, V.S.; LAO, L.L.; LIN-LIU, Y.R.; MILLER, R.L.; PETRIE, T.W.; POLITZER, P.A. et al.
Partner: UNT Libraries Government Documents Department

Modeling of electron cyclotron current drive experiments on DIII-D

Description: Electron Cyclotron Current Drive (ECCD) is considered a leading candidate for current profile control in Advanced Tokamak (AT) operation. Localized ECCD has been clearly demonstrated in recent proof-of-principle experiments on DIII-D. The measured ECCD efficiency near the magnetic axis agrees well with standard theoretical predictions. However, for off-axis current drive the normalized experimental efficiency does not decrease with minor radius as expected from the standard theory; the observed reduction of ECCD efficiency due to trapped electron effects in the off-axis cases is smaller than theoretical predictions. The standard approach of modeling ECCD in tokamaks has been based on the bounce-average calculations, which assume the bounce frequency is much larger than the effective collision frequency for trapped electrons at all energies. The assumption is clearly invalid at low energies. Finite collisionality will effectively reduce the trapped electron fraction, hence, increase current drive efficiency. Here, a velocity-space connection formula is proposed to estimate the collisionality effect on electron cyclotron current drive efficiency. The collisionality correction gives modest improvement in agreement between theoretical and recent DIII-D experimental results.
Date: May 1, 1999
Creator: Lin-Liu, Y.R.; Chan, V.S.; Luce, T.C.; Prater, R.; Sauter, O. & Harvey, R.W.
Partner: UNT Libraries Government Documents Department

Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

Description: The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.
Date: October 7, 2003
Creator: Jardin, S.C.; Kessel, C.E.; Mau, T.K.; Miller, R.L.; Najmabadi, F.; Chan, V.S. et al.
Partner: UNT Libraries Government Documents Department

Stability in High Gain Plasmas in DIII-D

Description: Fusion power gain has been increased by a factor of 3 in DIII-D plasmas through the use of strong discharge shaping and tailoring of the pressure and current density profiles. H-mode plasmas with weak or negative central magnetic shear are found to have neoclassical ion confinement throughout most of the plasma volume. Improved MHD stability is achieved by controlling the plasma pressure profile width. The highest fusion power gain Q (ratio of fusion power to input power) in deuterium plasmas was 0.0015. which extrapolates to an equivalent Q of 0.32 in a deuterium-tritium plasma and is similar to values achieved in tokamaks of larger size and magnetic fields.
Date: January 1, 1997
Creator: Lazarus, E. A.; Hong, R. M.; Navratil, G. A.; Sabbagh, S.; Strait, E. J.; Rice, B. W. et al.
Partner: UNT Libraries Government Documents Department

Simulation of enhanced tokamak performance on DIII-D using fast wave current drive

Description: The fast magnetosonic wave is now recognized to be a leading candidate for noninductive for the tokamak reactor due to the ability of the wave to penetrate to the hot dense core region. Fast wave current drive (FWCD) experiments on D3D have realized up to 120 kA of rf current drive, with up to 40% of the plasma current driven noninductively. The success of these experiments at 60 MHZ with a 2 MW transmitter source capability has led to a major upgrade of the FWCD system. Two additional transmitters, 30 to 120 NM, with a 2 MW source capability each, will be added together with two new four-strap antennas in early 1994. Another major thrust of the D3-D program is to develop advanced tokamak modes of operation, simultaneously demonstrating improvements in confinement and stability in quasi-steady-state operation. In some of the initial advanced tokamak experiments on D3-D with neutral beam heated (NBI) discharges it has been demonstrated that energy confinement nine can be improved by rapidly elongating the plasma to force the current density profile to be more centrally peaked. However, this high-l[sub i] phase of the discharge with the commensurate improvement in confinement is transient as the current density profile relaxes. By applying FWCD to the core of such a [kappa]-ramped discharge it may be possible to sustain the high internal inductance and elevated confinement.Using computational tools validated on the initial DM-D FWCD experiments we find that such a high-l[sub i] advanced tokamak discharge should be capable of sustainment at the 1 MA level with the upgraded capability of the FWCD system.
Date: May 1, 1993
Creator: deGrassie, J.S.; Lin-Liu, Y.R. Petty, C.C.; Pinsker, R.I.; Chan, V.S.; Prater, R.; St. John, H. (General Atomics, San Diego, CA (United States)) et al.
Partner: UNT Libraries Government Documents Department

Confinement and stability of VH-mode discharges in the DIII-D tokamak

Description: A regime of very high confinement (VH-mode) has been observed in neutral beam-heated deuterium discharges in the DIII-D tokamak with thermal energy confinement times up to [approx]3.6 times that predicted by the ITER-89P L-mode scaling and 2 times that predicted by ELM-free H-mode thermal confinement scalings. This high confinement has led to increased plasma performance, n[sub D] (0)T[sub i](0)[tau][sub E] = 2 [times] 10[sup 20] m[sup [minus]3] keV sec with I[sub p] = 1.6 MA, B[sub T] = 2.1 T, Z[sub eff] [le] 2. Detailed transport analysis shows a correspondence between the large decrease in thermal diffusivity in the region 0.75 [le] [rho] [le] 0.9 and the development of a strong shear in the radial electric field in the same region. This suggests that stabilization of turbulence by sheared E [times] B flow is responsible for the improved confinement in VH-mode. A substantial fraction of the edge plasma entering the second regime of stability may also contribute to the increase in confinement. The duration of the VH-mode phase has been lengthened by feedback controlling the input power to limit plasma beta.
Date: October 1, 1992
Creator: Taylor, T.S.; Osborne, T.H.; Burrell, K.H.; Carlstrom, T.N.; Chan, V.S.; Chu, M.S. et al.
Partner: UNT Libraries Government Documents Department

Regimes of improved confinement and stability in DIII-D obtained through current profile modifications

Description: Several regimes of improved confinement and stability have been obtained in recent experiments in the DIII-D tokamak by dynamically varying the toroidal current density profile to transiently produce a poloidal magnetic field profile with more favorable confinement and stability properties. A very peaked current density profile with high plasma internal inductance, [ell][sub i], is produced either by a rapid change in the plasma poloidal cross section or by a rapid change in the total plasma current. Values of thermal energy confinement times nearly 1.8 times the JET/DIII-D ELM-free H-mode thermal confinement scaling are obtained. The confinement enhancement factor over the ITER89-P L-mode confinement scaling, H, is as high as 3. Normalized toroidal beta, [beta][sub N], greater than 6%-m-T/MA and values of the product [beta][sub N]H greater than 15 have also been obtained. Both the confinement and the maximum achievable [beta] vary with [ell][sub i] and decrease as the current profile relaxes. For strongly shaped H-mode discharges, in addition to the current density profile peakedness, as measured by [ell][sub i] other current profile parameters, such as its distribution near the edge region, may also affect the confinement enhancement.
Date: September 1, 1992
Creator: Lao, L.L.; Ferron, J.R.; Taylor, T.S.; Chan, V.S.; Osborne, T.H.; Burrell, K.H. et al.
Partner: UNT Libraries Government Documents Department