N. A. Uckan

Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor plasma (Tokamak Ignition/Burn Experimental Reactor) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-dimensional transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability.

EBT Ring Physics

The purposes of the workshop were (1) to evaluate the status of the current experimental and theoretical understanding of the hot electron ring properties, (2) to identify the dominant physical processes that influence ring formation and scaling and their optimal behavior, and (3) to determine critical areas for near-term research.

Confinement capability of ITER-EDA design

Confinement capability of the ITER-EDA (R=7.75 m, I=25 MA) operational scenarios is evaluated and compared with the ITER CDA (R=6 m, 22 MA). The ignition capability of ITER EDA is somewhat higher than that of CDA by a factor of 1.1-1.2 with empirical power law scalings and by a factor of 1.5-2 with offset linear scalings. Simulations with the RLW /spl chi/(/spl nabla/T/sub e/)/sub crit/ model show that both the EDA and CDA scenarios operates in L-mode, however CDA ignition margin is much smaller. With empirical scalings, the required L-mode confinement enhancement factor [H=/spl tau//sub E///spl tau//sub E/(scaling)] corresponding to, for example, ITER89-P L-mode scaling would be 1.5-1.6 in ITER EDA relative to 1.8 in CDA for 10% He (plus 1% Be) concentration. At a higher concentration of He of 20-25%, the confinement capability is deteriorated and the required confinement enhancement factor (over empirical L-mode scalings) is /spl ges/2. The Ohmic confinement time is a factor of two higher in the EDA design (as compared to the CDA), yielding a strong reduction in the auxiliary power required to reach ignition. In 1.5-D simulations with L-mode enhancement factors of H/spl ges/1.2 allowed ohmic ignition with 25 MA, ignition was aided by initially peaked density profiles (and low He content) during the start-up.

ITER design: Physics basis for size, confinement capability power levels and burn control

The ITER-EDA(93) design has been analyzed to evaluate the physics basis for: (i) size and design trade-off issues, (ii) confinement capability, (iii) power levels, and (iv) burn control.

Safety Analyses for Transient Behavior of Plasma and In-vessel Components during Plasma Abnormal Events in Fusion Reactor

Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several para...

Analyses of passive plasma shutdown during ex-vessel loss of coolant accident in the first wall/shield blanket of fusion reactor

An ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of a fusion reactor has been analyzed by a hybrid code consisting of plasma dynamics and heat transfer analysis of in-vessel components. We investigated possibility of passive plasma shutdown scenario during the accident in International Thermonuclear Experimental Reactor (ITER). The safety analysis code which we developed can treat impurity concentration from the first wall and the divertor with a transport probability into the main plasma and a time delay given as input. It was found that the plasma is passively shutdown by a density limit disruption due to beryllium release from heated first wall surfaces about 168 seconds after the LOCA, when the transport probability of beryllium from the first wall into the plasma and the time delay were assumed to be 10−2 and the energy confinement time, respectively. At that time, the surface temperature of the outboard center (plasma facing component (PFC) with beryllium) and the temper...

Physics Guidelines for the Compact Ignition Tokamak

The goal of the Compact Ignition Tokamak (CIT)d program is to provide a cost-effective route to the production of a burning deuterium-tritium plasma, so that alpha-particle effects may be studied. A key issue to be studied in the CIT is whether alpha power behaves like other power sources in affecting tokamak plasma confinement. The program is managed by the Princeton Physics Laboratory and includes broad community involvement. Guidelines for the preliminary design effort have been provided by the Ignition Technical Oversight Committee in discussion with the tokamak community. The reference design is a tokamak with a high filed (10 T), high current (10 MA), poloidal divertor, and liquid-nitrogen-cooled coils. It is a small, high-power-density device of the type proposed by Bruno Coppi (MIT). It has a major radius of 1.23 m, a minor radius of 0.43 m, and plasma elipticity of 1.8. This paper reviews the aims of the program and the basis for the physics guidelines. The role of the CIT in the longer-term tokamak program is briefly discussed. 23 refs., 9 figs., 1 tab.

Development of safety assessment method for plasma anomaly events in fusion reactors

The objective of the study is to provide a safety assessment method for plasma transients including thermal response of in-vessel components. We developed a plasma physics model for safety analysis which has been implemented in a safety analysis code (SAFALY). The SAFALY code consists of a 0-D plasma dynamics model and a 1-D thermal behavior model of in-vessel components in the thickness direction. The code can treat hydraulic accidents using the results from a hydraulic code and analyze a passive plasma shutdown due to the impurity release from the wall. The overpower events in International Thermonuclear Experimental Reactor (ITER) were investigated, when the fueling rate and confinement improvement changes. The results show no significant damage to the confinement boundary of ITER is expected, as long as the cooling system works normally.

Transport analysis of ignited and current-driven ITER designs

The International Thermonuclear Experimental Reactor (ITER) design is intended to provide an engineering test of both ignited and current-driven operation. Both modes have been modeled using the WHIST transport code. Simulations indicate that tangential neutral beam injection inclined at a small angle to the horizontal midplane provides a means of controlling the heating and current drive profiles over a greater density range than near-normal injection, that a long current ramp of approx.100 s is needed to avoid skin currents during inductive startup, and that sawtooth activity is important in assessing volt-second consumption.

Assessment of a Compact Torsatron Reactor, ATFSR

Confinement and engineering issues of a small (average minor radius a-barapprox. =1 m) moderate-aspect-ratio torsatron reactor are evaluated. The Advanced Toroidal Facility design is used as a starting point because of its relatively low aspect ratio and high beta capabilities. The major limitation of the compact size is the lack of space under the helical coils for the blanket and shield. Some combination of lower aspect ratio coils, higher coil current density, thinner coils, and more effective shielding material under the coils should be incorporated into future designs to improve the feasibility of small torsatron reactor concepts. Current neoclassical confinement models for helically trapped particles show that a large radial electric field (in terms of the electric potential, e phi/Tgreater than or equal to3) is necessary to achieve ignition in a device of this size.