R.J. Hawryluk

Principal physics developments evaluated in the ITER design review

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

Development of ITER 15 MA ELMy H-mode inductive scenario

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation.

Importance of plasma response to nonaxisymmetric perturbations in tokamaks

Tokamaks are sensitive to deviations from axisymmetry as small as δB/B0∼10−4. These nonaxisymmetric perturbations greatly modify plasma confinement and performance by either destroying magnetic surfaces with subsequent locking or deforming magnetic surfaces with associated nonambipolar transport. The Ideal Perturbed Equilibrium Code (IPEC) calculates ideal perturbed equilibria and provides important basis for understanding the sensitivity of tokamak plasmas to perturbations. IPEC calculations indicate that the ideal plasma response, or equivalently the effect by ideally perturbed plasma currents, is essential to explain locking experiments on National Spherical Torus eXperiment (NSTX) and DIII-D. The ideal plasma response is also important for neoclassical toroidal viscosity (NTV) in nonambipolar transport. The consistency between NTV theory and magnetic braking experiments on NSTX and DIII-D can be improved when the variation in the field strength in IPEC is coupled with generalized NTV theory. These pla...

First observation of ELM pacing with vertical jogs in a spherical torus

Experiments in a number of conventional aspect ratio tokamaks have been successful in pacing edge localized modes (ELMs) by rapid vertical jogging of the plasma. This paper demonstrates the first pacing of ELMs in a spherical torus plasma. Applied 30 Hz vertical jogs synchronized the ELMs with the upward motion of the plasma. 45 Hz jogs also lead to an increase in the ELM frequency, though the synchronization of the ELMs and jogs was unclear. A reduction in the ELM energy was observed at the higher driven ELM frequencies.

23rd IAEA Fusion Energy Conference: summary of sessions EX/C and ICC

An overview is given of recent experimental results in the areas of innovative confinement concepts, operational scenarios and confinement experiments as presented at the 2010 IAEA Fusion Energy Conference. Important new findings are presented from fusion devices worldwide, with a strong focus towards the scientific and technical issues associated with ITER and W7-X devices, presently under construction.

Report of the FESAC Panel on a Burning Plasma Program Strategy to Advance Fusion Energy

This panel was set up by the U.S. Department of Energys Fusion Energy Sciences Advisory Committee in response to a request from the department to prepare a strategy for the study of burning fusion plasmas. Experimental study of a burning plasma has long been a goal of the U.S. science-based fusion energy program. There is an overwhelming consensus among fusion scientists that we are now ready scientifically, and have the full technical capability, to embark on this step. The fusion community is prepared to construct a facility that will allow us to produce this new plasma state in the laboratory, uncover the new physics associated with the fusion burn, and develop and test new technology essential for fusion power. Given this background, the panel has produced a strategy to enable the United States to proceed with this crucial next step in fusion energy science. The strategy was constructed with awareness that the burning plasma program is only one major component in a comprehensive development plan for fusion energy. A strong core science and technology program focused on fundamental understanding, confinement configuration optimization, and the development of plasma and fusion technologies essential to the realization of fusion energy. The core program will also be essential to the successful guidance and exploitation of the burning plasma program, providing the necessary knowledge base and scientific workforce.

Recommendations on the Nature and Level of U.S. Participation in the International Thermonuclear Experimental Reactor (ITER) Extension of the Engineering Design Activities

The Department of Energy (DOE) Office of Energy Research chartered through the Fusion Energy Sciences Advisory Committee (FESAC) a panel to “address the topic of U.S. participation in an ITER construction phase, assuming the ITER Parties decide to proceed with construction”. Given that there is expected to be a transition period of 3 to 5 years between the conclusion of the Engineering Design Activities (EDA) and the possible construction start, the DOE Office of Energy Research expanded the charge to “include the U.S. role in an interim period between the EDA and construction”.This panel has heard presentations and received input from a wide cross-section of parties with an interest in the fusion program. The panel concluded it could best fulfill its responsibility under this charge by considering the fusion energy science and technology portion of the U.S. program in its entirely. Accordingly, the panel is making some recommendations for optimum use of the transition period considering the goals of the fusion program and budget pressures.

High poloidal beta equilibria in TFTR limited by a natural inboard poloidal field null

Recent operation of the Tokamak Fusion Test Reactor TFTR, has produced plasma equilibria with values of {Lambda} {triple bond} {beta}{sub p eq} + l{sub i}/2 as large as 7, {epsilon}{beta}{sub p dia} {triple bond} 2{mu}{sub 0}{epsilon}/{much lt}B{sub p}{much gt}{sup 2} as large as 1.6, and Troyon normalized diamagnetic beta, {beta}{sub N dia} {triple bond} 10{sup 8} aB{sub 0}/I{sub p} as large as 4.7. When {epsilon}{beta}{sub p dia} {approx gt} 1.25, a separatrix entered the vacuum chamber, producing a naturally diverted discharge which was sustained for many energy confinement times, {tau}{sub E}. The largest values of {epsilon}{beta}{sub p} and plasma stored energy were obtained when the plasma current was ramped down prior to neutral beam injection. The measured peak ion and electron temperatures were as large as 24 keV and 8.5 keV, respectively. Plasma stored energy in excess of 2.5 MJ and {tau}{sub E} greater than 130 msec were obtained. Confinement times of greater than 3 times that expected from L-mode predictions have been achieved. The fusion power gain. Q{sub DD}, reached a values of 1.3 {times} 10{sup {minus}3} in a discharge with I{sub p} = 1 MA and {epsilon}{beta}{sub p dia} = 0.85. A large, sustained negative loop voltage duringmorexa0» the steady state portion of the discharge indicates that a substantial non-inductive component of I{sub p} exists in these plasmas. Transport code analysis indicates that the bootstrap current constitutes up to 65% of I{sup p}. Magnetohydrodynamic (MHD) ballooning stability analysis shows that while these plasmas are near, or at the {beta}{sub p} limit, the pressure gradient in the plasma core is in the first region of stability to high-n modes. 24 refs., 10 figs.«xa0less