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Compact DT fusion spherical tori at modest fields

Description: A spherical torus is obtained by retaining only the indispensable components on the inboard side of a tokamak plasma, such as a cooled, normal conductor that carries current to produce a toroidal magnetic field. The resulting device features an exceptionally small aspect ratio (typically 2-to-1 elongation), and ramp-up and maintenance of the plasma current primarily by noninductive means. The tokamak plasma takes on the spherical shape with a modest hole through the center, suggesting the name of spherical torus. This paper reviews the initial assessments of near-term DT fusion devices based on the spherical torus concept.
Date: January 1, 1985
Creator: Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Tn/tau//sub E/ comparisons of tokamak reactors of different aspect ratios

Description: The ignition parameter Tn/tau//sub E/ of the International Tokamak Experimental Reactor (ITER) with an aspect ratio A = 2.6, a major radius R = 5.8 m, a plasma current I/sub p/ = 22 MA, and an external toroidal field of B/sub t0/ = 5 T at R; a High Field Experimental Reactor (HFER) with A = 6.0, R = 6.0 m, I/sub p/ = 8.2 MA, and B/sub t0/ = 12.4 T; and a Spherical Torus Experimental Reactor (STER) with A = 1.2, R = 1.5 m, I/sub p/ = 22 MA, and B/sub t0/ = 0.88 T are compared using a set of presently popular empirical or semi-empirical confinement scaling expressions. Since the physics basis for these scaling expressions are not determined, the results of comparison are best suited to helping identify issues critical to furthering our understanding of tokamak confinement. In the case where the plasma pressure is limited by the Troyon beta scaling using B/sub t0/, the scaling expressions with relatively strong A-dependence, favor the HFER, while those with relatively strong I/sub p/-dependence favor the STER or suggest comparable Tn/tau//sub E/. In the case where the plasma pressure is determined by a second stability /epsilon//beta//sub p/ limit (/epsilon/ = 1/A), the projections of the T-10 Goldston, Odajima-Shimamura, and Rebut-Lallia scaling expressions are also compared. There it is found that the above STER and HFER would suffice if /epsilon//beta//sub p/ = 0.5. These varied projections highlight the uncertain I/sub p/, A and R dependencies in empirical tokamak confinement scaling, which are likely to be resolved only by tests at very low and high A. The results also indicate that whether the plasma pressure is determined by a high /epsilon//beta//sub p/ limit is critical to the viability of the HFER and the STER compared to ITER, even if the ...
Date: January 1, 1989
Creator: Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Spherical torus, compact fusion at low field

Description: A spherical torus is obtained by retaining only the indispensable components on the inboard side of a tokamak plasma, such as a cooled, normal conductor that carries current to produce a toroidal magnetic field. The resulting device features an exceptionally small aspect ratio (ranging from below 2 to about 1.3), a naturally elongated D-shaped plasma cross section, and ramp-up of the plasma current primarily by noninductive means. As a result of the favorable dependence of the tokamak plasma behavior to decreasing aspect ratio, a spherical torus is projected to have small size, high beta, and modest field. Assuming Mirnov confinement scaling, an ignition spherical torus at a field of 2 T features a major radius of 1.5 m, a minor radius of 1.0 m, a plasma current of 14 MA, comparable toroidal and poloidal field coil currents, an average beta of 24%, and a fusion power of 50 MW. At 2 T, a Q = 1 spherical torus will have a major radius of 0.8 m, a minor radius of 0.5 m, and a fusion power of a few megawatts.
Date: February 1, 1985
Creator: Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

CIT systems analysis and plasma engineering

Description: The report consists of viewgraphs. Topics covered are: development of optional pf coil configuration, disruption current quench simulation and impact on vacuum vessel design, and ''second opinion'' systems analysis of CIT design evolution.
Date: January 1, 1987
Creator: Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Equilibrium field coils and free-boundary equilibrium considerations for TNS

Description: The Next Step (TNS) tokamak is expected to have a D-shaped plasma to permit MHD stable operation with volume averaged-beta, anti..beta.. up to 10 percent. By following a procedure similar to the method of virtual casing, external coil arrangements were produced that reproduce the ideal vacuum vertical field B/sub z//sup ID/ for a D-shaped plasma to within a few percent root-mean-squared deviation. A typical coil system and a free-boundary equilibrium are shown. It is seen that D-shaped equilibria can be with the external coils. Again, three sets of coils are sufficient for centering and shaping the plasma for the range of anti..beta.. values of interest. It is concluded that placing the equilibrium field coils external to the toroidal field coils has the potential of substantially reducing the cost and complexity of the D-shaped TNS device. (MHR)
Date: January 1, 1977
Creator: Peng, Y.K.M. & Strickler, D.J.
Partner: UNT Libraries Government Documents Department

Two-point model for divertor transport

Description: Plasma transport along divertor field lines was investigated using a two-point model. This treatment requires considerably less effort to find solutions to the transport equations than previously used one-dimensional (1-D) models and is useful for studying general trends. It also can be a valuable tool for benchmarking more sophisticated models. The model was used to investigate the possibility of operating in the so-called high density, low temperature regime.
Date: April 1, 1984
Creator: Galambos, J.D. & Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Poloidal field-coil system of the fusion-engineering device

Description: The Fusion Engineering Design Center (FEDC) initiated a program in FY 81 directed towards the development of a Fusion Engineering Device (FED) tokamak design description. During the period from October 1980 to March 1981, the emphasis was on trade and design studies, in an effort to establish a baseline concept for the FED. This was followed by a period extending through September 1981 during which the chosen concept was examined in detail, and substantial progress was made towards a self-consistent FED design. The purpose of this paper is to describe the evolution of the poloidal field configuration in this design process, including the choice of an equilibrium field (EF) coil concept, and the operating scenario of a particular coil set based on that concept.
Date: January 1, 1981
Creator: Strickler, D.J. & Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

FED-A, an advanced performance FED based on low safety factor and current drive

Description: The FED-A study aims to quantify the potential improvement in cost-effectiveness of the Fusion Engineering Device (FED) by assuming low safety factor q (less than 2 as opposed to about 3) at the plasma edge and noninductive current drive (as opposed to only inductive current drive). The FED-A performance objectives are set to be : (1) ignition assuming International Tokamak Reactor (INTOR) plamsa confinement scaling, but still achieving a fusion power amplification Q greater than or equal to 5 when the confinement is degraded by a factor of 2; (2) neutron wall loading of about 1 MW/m/sup 2/, with 0.5 MW/m/sup 2/ as a conservative lower bound; and (3) more clearly power-reactor-like operations, such as steady state.
Date: August 1, 1983
Creator: Peng, Y.K.M. & Rutherford, P.H.
Partner: UNT Libraries Government Documents Department

Features of spherical torus plasmas

Description: The spherical torus is a very small aspect ratio (A < 2) confinement concept obtained by retaining only the indispensable components inboard to the plasma torus. MHD equilibrium calculations show that spherical torus plasmas with safety factor q > 2 are characterized by high toroidal beta (..beta../sub t/ > 0.2), low poloidal beta (..beta../sub p/ < 0.3), naturally large elongation (kappa greater than or equal to 2), large plasma current with I/sub p//(aB/sub t0/) up to about 7 MA/mT, strong paramagnetism (B/sub t//B/sub t0/ > 1.5), and strong plasma helicity (F comparable to THETA). A large near-omnigeneous region is seen at the large-major-radius, bad-curvature region of the plasma in comparison with the conventional tokamaks. These features combine to engender the spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost. Because of its strong paramagnetism and helicity, the spherical torus plasma shares some of the desirable features of spheromak and reversed-field pinch (RFP) plasmas, but with tokamak-like confinement and safety factor q. The general class of spherical tori, which includes the spherical tokamak (q > 1), the spherical pinch (1 > q > O), and the spherical RFP (q < O), have magnetic field configurations unique in comparison with conventional tokamaks and RFPs. 22 refs., 12 figs.
Date: December 1, 1985
Creator: Peng, Y.K.M. & Strickler, D.J.
Partner: UNT Libraries Government Documents Department

Parametric assessments of current rampup by Lower Hybrid Current Drive in TFCX

Description: A lower hybrid current drive code has been developed from two previous existing codes which incorporates the eikonal (WKB) approximation in toroidal canonical variables and an analytic treatment of quasi-linear absorption. Ray-tracing is performed for a 5 to 9 ray spectrum in a nonuniform, circular plasma with the local warm plasma dispersion relation. A parameter study in n/sub e/, T/sub e/, n/sub parallel/, and frequency is being carried out for two startup scenarios (i.e., expanding radius at the outboard midplane and full radius startup on the toroidal axis) for the superconducting toroidal field coil TFCX device. Driven current profiles and current drive efficiencies for both a Maxwellian background plasma and an electron distribution with an rf-enhanced tail are presented.
Date: January 1, 1984
Creator: Freije, S.A. & Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Ignition and burn criteria for D/sup 3/He tokamak and spherical torus reactors

Description: D-/sup 3/He ignition and burn criteria for tokamaks and spherical torus reactors are examined in a global analysis with profile corrections. Particle confinement and ash buildup effects are included with the power balance, which results in an increased sensitivity of the ignition criteria to losses via brehmsstrahlung and synchrotron radiation. Plasma beta scaling via an /epsilon//beta//sub p/ limit provides the needed aspect ratio (A) dependence, and permits an analysis in all A values of the first and second stability regimes. Energy confinement time (/tau//sub E/) associated with particle diffusion (/tau//sub p/) and energy conduction (/tau//sub c/) are used; parabolic profile are assumed with exponents /alpha//sub n/ = 1.0 and /alpha//sub T/ = 1.5; and we set /tau//sub p/ = 2/tau//sub c/. The ignition condition for minimum n/tau//sub E/ is found to be sensitive to /beta/ but not to the magnetic field. Steady state burn in second stability tokamaks (/epsilon//beta//sub p/ /ge/ 0.6) at high A (>4) with average synchrotron wall reflectivities below 95% requires n/tau//sub E/ above 5 /times/ 10/sup 21/ m/sup /minus/3/s or strong plasma elongation (/kappa/ > 3). Ignition in a spherical tori can be achieved with wall reflectivities below 80% and at n/tau//sub E/ /le/ 10/sup 21/ m/sup /minus/3/ s, without requiring strong plasma shaping or /epsilon//beta//sub p/ > 0.6. The need to minimize n/tau//sub E/ for ignition and burn strongly limits the synchrotron radiation loss to less than 20% of the fusion power for all values of A. Synchrotron power fractions can be increased to 40% by increasing n/tau//sub E/ to its upperbound of ignition. Further increases of this fraction can be obtained only by assuming preferential ash removal. 19 refs., 8 figs.
Date: January 1, 1989
Creator: Galambos, J.D. & Peng, Y-K. M.
Partner: UNT Libraries Government Documents Department

System studies for quasi-steady-state advanced physics tokamak

Description: Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated.
Date: November 1, 1983
Creator: Reid, R.L. & Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Advanced performance fusion engineering device based on low safety factor and current drive (FED-A)

Description: The FED-A study aims to quantify the potential improvement in cost-effectiveness of the Fusion Engineering Device (FED) by assuming low safety factor q at the plasma edge and noninductive current drive. The FED-A performance objectives (ignition, neutron wall load, and power-reactor-like operation) are set to be equal to or better than those of the FED Baseline. The results show that assuming magnetohydrodynamic (MHD) q/sub psi/ (edge) to be 1.8 permits reduction in device size and plasma current and leads to a 30% reduction in direct cost. A closely fitted, 1.5-cm-thick, continuous water-cooled shell made of the copper alloy AMAX-MZC (0.6 Cr, 0.1 Zr, 0.03 Mg) is proposed to provide a 0.5-s time constant, to help avoid disruption when q/sub psi/ passes near 2, and to mitigate disruption impact. The lower hybrid wave current drive in a cyclic density operation is proposed to achieve a quasi-steady-state operation permitting a design with low toroidal loop voltage and a 1000-s burn time.
Date: January 1, 1983
Creator: Peng, Y.K.M. & Rutherford, P.H.
Partner: UNT Libraries Government Documents Department

Topology of tokamak orbits

Description: The topology of all contained tokamak guiding center orbits is displayed in a three-dimensional constants-of-motion space. The treatment is perfectly general and holds for arbitrary axisymmetric MHD equilibria. We show that significant topological changes occur in the high-anti BETA (approximately &gt; 6%) cases which are associated with a region of absolute minimum B in the plasma.
Date: January 1, 1978
Creator: Rome, J. A. & Peng, Y. K.M.
Partner: UNT Libraries Government Documents Department

Topology of tokamak orbits

Description: Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having ..beta..'s of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher ..beta.. (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower ..beta.. case. The differences indicate the confinement of additional high energy (v ..-->.. c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well.
Date: September 1, 1978
Creator: Rome, J.A. & Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Very small aspect ratio tokamaks

Description: Magnetohydrodynamic (MHD) stability analyses are carried out for tokamak equilibria with A = 2.0 and 1.5 q(boundary)/q(axis) approximately equal to 2, and ..beta../sub p/ approximately equal to A/2. When a conducting wall is at a distance 0.2a from the plasma with A = 1.5, the critical anti ..beta../sub t/(= 8..pi..p/B/sub to/) values are found to be 0.47, 0.37, and 0.33 for toroidal mode numbers N = 1, 2, and 3, respectively. A linear extrapolation in 1/N ..-->.. 0 results in a anti beta/sub tc/ of 0.28, which is consistent with the results based on analyses of large N ballooning modes on each flux surface. For the case of A = 2.0, anti ..beta../sub tc/ values on plasma parameters are quantified together with suggested approaches to produce the plasma via adiabatic compression.
Date: January 1, 1978
Creator: Peng, Y.K.M. & Dory, R.A.
Partner: UNT Libraries Government Documents Department

Equilibrium field coil concepts for INTOR

Description: Methods are presented for reducing ampere-turn requirements in the EF coil system. It is shown that coil currents in an EF coil system external to the toroidal field coils can be substantially reduced by relaxing the triangularity of a D-shaped plasma. Further reductions are realized through a hybrid EF coil system using both internal and external coils. Equilibrium field coils for a poloidally asymmetric, single-null INTOR configuration are presented. It is shown that the shape of field lines in the plasma scrapeoff region and divertor channel improves as triangularity is reduced, but it does so at the possible expense of achievable stable beta values.
Date: August 1, 1981
Creator: Strickler, D.J.; Peng, Y.K.M. & Brown, T.G.
Partner: UNT Libraries Government Documents Department

MHD equilibrium methods for ITER (International Thermonuclear Experimental Reactor) PF (poloidal field) coil design and systems analysis

Description: Two versions of the Fusion Engineering Design Center (FEDC) free-boundary equilibrium code designed to computer the poloidal field (PF) coil current distribution of elongated, magnetically limited tokamak plasmas are demonstrated and applied to the systems analysis of the impact of plasma elongation on the design point of the International Thermonuclear Experimental Reactor (ITER). These notes were presented at the ITER Specialists' Meeting on the PF Coil System and Operational Scenario, held at the Max Planck Institute for Plasma Physics in Garching, Federal Republic of Germany, May 24--27, 1988. 8 refs., 6 figs., 4 tabs.
Date: March 1, 1989
Creator: Strickler, D.J.; Galambos, J.D. & Peng, Y.K.M.
Partner: UNT Libraries Government Documents Department

Prospects and status of low-aspect-ratio tokamaks

Description: The prospects for the low-aspect-ratio (A) tokamak to fulfill the requirements of viable fusion power plants are considered relative to the present status in data and modeling. Desirable physics and design features for an attractive Blanket Test Facility and power reactors are estimated for low-A tokamaks based on calculations improved with the latest data from small pioneering experiments. While these experiments have confirmed some of the recent predictions for low-A, they also identify the remaining issues that require verification before reliable projections can be made for these deuterium-tritium applications. The results show that the low-A regime of small size, modest field, and high current offers a path complementary to the standard and high A tokamaks in developing the full potential of fusion power.
Date: December 1994
Creator: Peng, Y. K. M.
Partner: UNT Libraries Government Documents Department

Science and Technology of the 10-MA Spherical Tori

Description: The Spherical Torus (ST) configuration has recently emerged as an example of confinement concept innovation that enables attractive steps in the development of fusion energy. The scientific potential for the ST has been indicated by recent encouraging results from START,2 CDX-U, and HIT. The scientific principles for the D-fueled ST will soon be tested by NSTX (National Spherical Torus Experiment3) in the U.S. and MAST (Mega-Amp Spherical Tokamak4) in the U.K. at the level of l-2 MA in plasma current. More recently, interest has grown in the U.S. in the possibility of near-term ST fusion burn devices at the level of 10 MA in plasma current. The missions for these devices would be to test burning plasma performance in a small, pulsed D-T-fueled ST (i.e., DTST) and to develop fusion energy technologies in a small steady state ST-based Volume Neutron Source (VNS). This paper reports the results of analysis of the key science and technology issues for these devices.
Date: November 14, 1999
Creator: Peng, Y-K.M.
Partner: UNT Libraries Government Documents Department

Hybrid equilibrium field coils for the ORNL TNS

Description: In this study, we make a comparative study of the power supplies required by interior and exterior (to the toroidal field (TF) coils) equilibrium field coils that are separately appropriate for high-..beta.., D-shaped plasmas in TNS. It is shown that the interior coils need power supplies that are an order of magnitude below those required by the exterior coils (while the latter case is much less difficult to build than the former). A hybrid EF coil concept is proposed that combines the interior and the exterior coils to retain their advantages in avoiding large interior coils while lowering the power supplied to the exterior coils by an order of magnitude.
Date: January 1, 1977
Creator: Peng, Y.K.M.; Strickler, D.J & Dory, R.A.
Partner: UNT Libraries Government Documents Department