R.H. Bulmer

Analysis of geometric variations in high-power tokamak divertors

Quantitative assessment of the performance of high-power tokamak divertors for a range of geometric variations is conducted using the MHD code Corsica (Crotinger et al 1997 Technical Report LLNL) and edge transport code UEDGE (Rognlien et al 1992 J. Nucl. Mater. 196–198 347). In a multi-parametric study the divertor performance is compared for a high-power tokamak with standard and snowflake (Ryutov 2007 Phys. Plasmas 14 064502) configurations for the same core plasma parameters. Divertor and edge quantities that are varied include x-point flux expansion, shape of target plates, and radiating impurity species and concentrations. For a range of studied cases, in the snowflake the peak heat load on the target plates is significantly reduced compared with the standard divertor due to larger plasma-wetted area and a larger fraction of power radiated in the edge.

The super X divertor (SXD) and a compact fusion neutron source (CFNS)

A new magnetic geometry, the super X divertor (SXD), is invented to solve severe heat exhaust problems in high power density fusion plasmas. SXD divertor plates are moved to the largest major radii inside the TF coils, increasing the wetted area by 2–3 and the line length by 2–5. Two-dimensional fluid simulations with SOLPS (Schneider et al 2006 SOLPS 2-D edge calculation code Contrib. Plasma Phys. 46) show a several-fold decrease in divertor heat flux and plasma temperature at the plate. A small high power density tokamak using SXD is proposed, for either (1) useful fusion applications using conservative physics, such as a component test facility (CTF) or fission–fusion hybrid, or (2) to develop more advanced physics modes for a pure fusion reactor in an integrated fusion environment.

TPX superconducting tokamak magnet system 1995 design and status overview

The TPX magnet preliminary design effort is summarized. Key results and accomplishments during preliminary design and supporting R&D are discussed, including conductor development, quench detection, TF and PF magnet design, conductor bending and forming, reaction heat treating, helium stubs, and winding pack insulation.

Modelling of magnetic perturbations and confinement in SSPX spheromak

The effects of magnetic perturbation on the quality of confinement in the sustained spheromak plasma experiment (SSPX) (Hooper E B et al 1999 Nucl. Fusion 39 863) are studied. The magnetic field is taken as a sum of an unperturbed axisymmetric magnetic field, B 0 , calculated from a magneto-hydrodynamic equilibrium, and a non-axisymmetric part, δB, representing the effects of kinking perturbation of the central plasma column. Poincare maps show that flux surfaces are strongly affected by perturbations with δB/B 0 ? 5%, with islands and stochastic regions covering a substantial part of plasma volume. The effective heat transport in the perturbed magnetic field, calculated by a Monte-Carlo procedure, is found to be consistent with the experimentally measured level of confinement.

Control of first-wall surface conditions in the 2XIIB magnetic mirror plasma confinement experiment

The control of first-wall surface conditions in the 2XIIB Magnetic Mirror Plasma Confinement experiment is described. Before each plasma shot, the first wall is covered with a freshly gettered titanium surface. Up to 5 MW of neutral beam power has been injected into 2XIIB, resulting in first-wall bombardment fluxes of 10

CORSICA modelling of ITER hybrid operation scenarios

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Observation of a pressure limit in a gun-driven spheromak

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Wall Conditioning and Power Balance for Spheromak Plasmas in SSPX

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Suppression of MHD fluctuations leading to improved confinement in a gun-driven spheromak.

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New mode of operating a magnetized coaxial plasma gun for injecting magnetic helicity into a spheromak.

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