The intrinsic properties of low temperature helium exhibit strong variations with changes in pressure and temperature. As a result, important design parameters such as the energy margin of superconductors in the cable in conduit configuration is a strong function of the initial operating conditions. Since the energy margin of CICCs is dominated by the available helium enthalpy, the dependence of the energy margin on initial conditions can be derived from an examination of the helium properties. For superconductors operating in the range of 6 to 8 K, the energy margin can be improved by a factor of three with a suitable choice of initial operating pressure. Other factors effecting the energy margin include the length of the initial heated zone, and the length of the heating pulse

The intrinsic properties of low temperature helium exhibit strong variations with changes in pressure and temperature. As a result, important design parameters such as the energy margin of superconductors in the cable in conduit configuration is a strong function of the initial operating conditions. Since the energy margin of CICC's is dominated by the available helium enthalpy, the dependence of the energy margin on initial conditions can be derived from an examination of the helium properties. For superconductors operating in the range of 6 to 8 K, the energy margin can be improved by a factor of three with a suitable choice of initial operating pressure. Other factors effecting the energy margin include the length of the initial heated zone, and the length of the heating pulse.

A set of universal scaling relations is presented describing the propagation of quench in CICC (cable-in-conduit superconducting magnet) magnets. Four distinct types of behavior are possible depending upon the length of the coil and the magnitude of the quench induced pressure rise. The boundaries separating these regions can be simply expressed in terms of L/sub q/ the initial quench length, and J the stabilizer current density, the two parameters likely to vary during standard operation. The phenomenon of thermal hydraulic quenchback (THQB) is also considered. It is shown that the conditions for the onset of THQB can also be cast as a set of universal scaling relations and easily superimposed on the quench diagram.< >

The Toroidal Field (TF) and the Central Solenoid (CS) coils of ITER experience a variety of transient heating loads. During normal operation, the conductor in the TF coil experiences nuclear heating, whose magnitude is a function of the plasma operating conditions. In addition, the TF coils experience minor heating due to AC losses and the friction between the support plates. The heating in the CS coil is mainly due to AC losses which is significant during plasma start-up and shut-down. Due to the pulsed operation, and because of the long length of the flow paths (similar to 800 meters), steady state conditions are not established during a single pulse. In order to determine the operating margins, time dependent analyses of both the TF and the CS coils are performed under various heating conditions and over many cycles. Quench detection methods based on flow measurements at the inlet and outlet of the flow paths have to distinguish between ''normal'' and quench related flows. Extensive analyses of the flow signals under different operating and quench conditions show that, with the proper differencing the signal due to quench can be extracted.

The GEM test coil, will be wound from 70 m of conductor identical to that used in the full scale magnet. The coil configuration will duplicate the field distribution of the full scale magnet and current control will duplicate full scale current decay characteristics. Therefore, quench/protection analysis of this coil will reveal very important information about the behavior of the full scale model. Due to the uncertainty associated with the contest between the cable, the conduit and the sheath, a parametric analysis has been performed in order to determine and bracket the behavior. With no electrical contact the quench evolves normally until, due to heat transfer from the sheath into the cable, the superconductor temperature becomes critical and the entire length becomes normal. >

Tokamak designs based on copper and copper alloy magnets could be used for the investigation of the physics issues associated with long pulse (>40 sec) ignited operation during the next phase of fusion research. The engineering characteristics of designs with magnets that use copper or beryllium copper alloys are presented. Active cooling of the magnets with either liquid nitrogen or water is considered. Inertial cooling is also discussed. The physics performance of the designs is calculated and compared to the performance of the ''Physics Phase'' ITER design, and the BPX tokamak.

This paper presents a theoretical model describing quench propagation in cable in conduit conductors (CICC) with an additional central flow channel. The central channel is used to enhance the flow capabilities in the conduit during steady state operation as well as during quench events. Such a system is the proposed design for certain conductors in the International Thermonuclear Experimental Reactor (ITER). Here, the additional channel is formed by a metal spring located at the center of the conduit. We describe the separate thermal evolution in both the cable bundle and the central channel; in particular, the mass, momentum and heat transfer due to flow between the cable bundle and the central channel are included in the model. Several simplifications are introduced which greatly reduce the complexity of the model without sacrificing accuracy. The resulting reduced model is solved both numerically and approximately analytically for ITER parameters

In D-T and D-3He plasmas, ICRF heating at the second harmonic of deuterium results in a modification to the distribution function of the heated ions. The authors describe the results of such effects on the dynamic behaviour of plasmas. Using a 0-D plasma transport model, the effect of ICRF heating on an operating point is analysed for ITER and for a D-3He tokamak. To describe the dynamic behaviour, a model for the characteristic time of tail relaxation, τξ, is developed and a feedback model based on auxiliary power is presented. The stabilization of temperature perturbations in a D-3He plasma is simulated for various values of τξ

Engineering aspects of tokamak designs which could address the physics of long-pulse hydrogen/deuterium operation and alpha-particle-dominated heating (Q>5) are presented. The main characteristics of these designs are their relatively small size, their resistive toroidal field coils, and their long pulse capability. In D-T operation this machine could satisfy the basic physics requirements for the ITER (International Thermonuclear Experimental Reactor) physics mission at a much lower cost than the present superconducting ITER design. For these designs, demountable TF (toroidal field) coils similar to Alcator C-MOD are considered, and steady-state water cooling of the magnets is investigated. The demountable nature of the TF magnet simplifies the assembly and maintenance operations. Internal poloidal field coils minimize this system and increase its flexibility.<>

A new code, named SUPERCODE, has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of their physics and engineering models. The latter contain state-of-the-art plasma physics modeling but are limited in engineering content and are time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2-D plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formulation there is only a modest increase in CPU time over 0-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, we have expended considerable effort to make the code user- and programmer-friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community.

The presence of cyclic temperature fluctuations in the plasma, driven by either sawteeth effects, marfes or other non-classical phenomena muddles the process to control the thermal instability. Furthermore, the power fluctuations corresponding to the changes in temperature profiles could adversely affect the lifetime of components in reactors due to thermal cycling. Methods for minimizing the effects, both on the thermal instability control method and the fusion power variations, are discussed. Detailed 11 D calculations are performed, and the results are justified using simplified models.

A 0-D transport model is used as a means of examining the Compact Ignition Tokamak (CIT) thermal instability. It is shown that auxiliary power modulation can be used for mapping the tokamak operating space, choosing the operating point, and for thermal stability control. Stable operation with Q~65 can be obtained. The effect of the delay time τd of the feedback system is investigated, and the tradeoffs between the τd and the maximum allowable temperature fluctuation are determined

In this paper we write the 0-D energy balance in terms of the parameters , and . The minimum value of W required to achieve ohmic ignition (i.e. II = 0) in the generic case (i.e. no n or T dependance on τ), and for various dependencies of τ on n and T is W ∼ 15 in all cases. Also the value of W below which no access, even with the aid of auxiliary power, to high temperature alpha dominated operation is possible is W ∼ 2 in all cases. Optimized designs are obtained by minimizing the ratio of the auxiliary power pa to the ohmic power pΩ at the Cordey pass. The optimization is consistent with the physics requirements, the toroidal field coil volume and stress constraints, and with the volt-second and stress constraints of the ohmic transformer. Results under various τ dependencies are presented.