Validation of 1-D transport and sawtooth models for ITER

In this paper the authors describe progress on validating a number of local transport models by comparing their predictions with relevant experimental data from a range of tokamaks in the ITER profile database. This database, the testing procedure and results are discussed. In addition a model for sawtooth oscillations is used to investigate their effect in an ITER plasma with alpha-particles

Continuous pellet fueling experiments on D-III

A centrifuge pellet injector developed at ORNL was used to continuously fuel beam-heated limiter discharges in D-III. This injector was capable of producing and maintaining a high density neutral beam-heated plasma without auxilary gas fueling. Viewgraphs from the presentation are included

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

This document is one of four describing studies performed in FY 1982 within the context of the Fusion Engineering Device (FED) Program for the Office of Fusion Energy, U.S. Department of Energy. The documents are: 1. FED Baseline Engineering Studies (ORNL/FEDC-82/2), 2. FED-A, An Advanced Performance FED Based on Low Safety Factor and Current Drive (this document), 3. FED-R, A Fusion Device Utilizing Resistive Magnets (ORNL/FEDC-82/1), and 4. Technology Demonstration Facility TDF. These studies extend the FED Baseline concept of FY 1981 and develop innovative and alternative concepts for the FED. The FED-A study project was carried out as part of the Innovative and Alternative Tokamak FED studies, under the direction of P. H. Rutherford, which were part of the national FED program during FY 1982. The studies were performed jointly by senior scientists in the magnetic fusion community and the staff of the Fusion Engineering Design Center (FEDC). Y-K. M. Peng of the FEDC, on assignment from Oak Ridge National Laboratory, served as the design manager

Evaluation of current drive requirements and operating characteristics of a high bootstrap fraction advanced tokamak reactor

The reactor potential of some advanced physics operating modes proposed for the TPX physics program are examined. A moderate aspect ratio (A = 4.5 as in TPX), 2 GW reactor is analyzed because of its potential for steady-state, non-inductive operation with high bootstrap current fraction. Particle, energy and toroidal current equations are evolved to steady-state conditions using the 1-1/2-D time-dependent WHIST transport code. The solutions are therefore consistent with particle, energy and current sources and assumed transport models. Fast wave current drive (FWCD) provides the axial seed current. The bootstrap current typically provides 80-90% of the current, while feedback on the lower hybrid current drive (LHCD) power maintains the total current. The sensitivity of the plasma power amplification factor, Q {equivalent_to} P{sub fus}/P{sub aux}, to variations in the plasma properties is examined. The auxiliary current drive power, P{sub aux} = P{sub LH} + P{sub FW}; bootstrap current fraction: current drive efficiency; and other parameters are evaluated. The plasma is thermodynamically stable for the energy confinement model assumed (a multiple of ITER89P). The FWCD and LHCD sources provide attractive control possibilities, not only for the current profile, but also for the total fusion power since the gain on the incremental auxiliary power is typically 10-30 in these calculations when overall Q {approx} 30

Transport simulation of ITER (International Thermonuclear Engineering Reactor) startup

The present International Thermonuclear Engineering Reactor (ITER) reference configurations are the Technology Phase,'' in which the plasma current is maintained noninductively at a subignition density, and the Physics Phase,'' which is ignited but requires inductive maintenance of the current. The WHIST 1.5-D transport code is used to evaluate the volt-second requirements of both configurations. A slow current ramp (60-80's) is required for fixed-radius startup in ITER to avoid hollow current density profiles. To reach the operating point requires about 203 V{center dot}s for the Technology Phase (18 MA) and about 270 V{center dot}s for the Physics Phase (22 MA). The resistive losses can be reduced with expanding-radius startup. 5 refs., 4 figs

Neutral beam energy and power requirements for expanding radius and full bore startup of tokamak reactors

Natural beam power and energy requirements are compared for full density full bore and expanding radius startup scenarios in an elongated plasma, The Next Step (TNS), as a function of beam pulse time and plasma density. Because of the similarity of parameters, the results should also be applicable to Engineering Test Facility (ETF) and International Tokamak Reactor (INTOR) studies. A transport model consisting of neoclassical ion conduction and anomalous electron conduction and diffusion based on ALCATOR scaling leads to average densities in the range approx. 0.8 to 1.2 x 10/sup 14/ cm/sup -3/ being sufficient for ignition. Neutral deuterium beam energies in the range 120 to 180 keV are adequate for penetration, with the required power injected into the plasma decreasing with increasing beam energy. The neutral beam power decreases strongly with increasing beam pulse length b/sub b/ until t/sub b/ exceeds a few total energy confinement times, yielding b/sub b/ approx. = 4 to 6 s for the TNS plasma

Start-up scenarios for tokamak reactors

The effects of injecting cold DT ice pellets into a tokamak plasma in concert with neutral beam heating has been investigated for a full bore plasma start-up. A 1-D multifluid transport code has been used which incorporates a divertor/limiter model, neutral particle recycling, discrete pellet ablation physics, neutral beam, and alpha heating. Results show the effects of transport scaling laws used. We find that further study is required to ascertain the sign of the temperture gradient driven particle flux and heat flux due to microinstabilities residing in the central regions of the plasma where eta identical with delta lnT/delta lnN much greater than 1 (or possibly eta < 0). In addition preliminary results on scenarios using beam heating and neutral recycling for an expanding radius start-up are presented

Tritium pellet injection sequences for TFTR

Tritium pellet injection into neutral deuterium, beam heated deuterium plasmas in the Tokamak Fusion Test Reactor (TFTR) is shown to be an attractive means of (1) minimizing tritium use per tritium discharge and over a sequence of tritium discharges; (2) greatly reducing the tritium load in the walls, limiters, getters, and cryopanels; (3) maintaining or improving instantaneous neutron production (Q); (4) reducing or eliminating deuterium-tritium (D-T) neutron production in non-optimized discharges; and (5) generally adding flexibility to the experimental sequences leading to optimal Q operation. Transport analyses of both compression and full-bore TFTR plasmas are used to support the above observations and to provide the basis for a proposed eight-pellet gas gun injector for the 1986 tritium experiments