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An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fullymore » explored.« less

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). Its mission is to develop the physics understanding of the compact stellarator and evaluate its potential for future fusion energy systems. Compact stellarators use 3D plasma shaping to produce a magnetic configuration that can be steady state without current drive or feedback control of instabilities. The NCSX has major radius 1.4 m, aspect ratio 4.4, 3 field periods, and a quasi-axisymmetric magnetic field. It is predicted to be stable and have good magnetic surfaces at beta > 4% and to have tokamak-like confinement properties. The device will provide the plasma configuration flexibility and the heating and diagnostic access needed to test physics predictions. Component production has advanced substantially since the first contracts were placed in 2004. Manufacture of the vacuum vessel was completed in 2006. All eighteen modular coil winding forms have been delivered, and twelve modular coils have been wound and epoxy impregnated. A contract for the (planar) toroidal field coils was placed in 2006 and manufacture is in progress. Assembly activities have begun and will be the project's main focus in the next few years. The engineering challenge of NCSX is to meet the requirements for complex geometries and tight tolerances within the cost and schedule constraints of a construction project. This paper will focus on how the engineering challenges of component production have been resolved, and how the assembly challenges are being met.

A conceptual design has been developed for the recovery of un-expended fuel, ash, and associated post-detonation products from a ~ 2 GW IFE power reactor. The conceptual design incorporates systems for the safe, efficient collection, processing, and purification of IFE plasma exhaust fuel components. The system has been designed and sized such that tritium bred within blankets can also be collected, processed, and introduced into the fuel cycle. The system is nominally sized to process ~2 kg of tritium per day and is designed to link directly to the target chamber mechanical pumping system. The plasma exhaust can be directly processed from the exhaust of the vacuum pumping system or can be processed in batch mode from buffer vessels in the receiving and analysis system (RAS). Systems for the accurate measurement of material in-process (MIP) have been included. Design emphasis is on safety, reliability, redundancy, and efficiency in order to maximize availability. The primary goal of the fuel recovery system (FRS) design is to economically recycle components of IFE fuel back to the target manufacturers in a fashion by which fuel components are rapidly made available for re-use thus lowering the total active inventory. The FRS design is presented as a facility sub-system in the context of supporting the safe and efficient operation of the IFE target chamber.

A conceptual design for a magnetic intervention system is presented in support of a 2 GW IFE direct drive fusion power reactor. The system is designed employing a cusp field to deflect ions generated by an IFE implosion away from the first wall of the reactor core and into specifically designed ion dumps. The magnetic coil system will employ liquid helium cooled 5083 Aluminum alloy casing on a Rutherford NbTi cable. The cables are configured as four double pancakes with a 5083 Aluminum alloy case for structural support. The conceptual design and corresponding preliminary load and field calculations will be presented.

The Tokamak Fusion Test Reactor (TFTR) operated for 15 years, from 1982 to 1997. From November 1993 to April 1997 a mixture of deuterium-tritium (D-T) was used to fuel experiments, leaving the reactor system in an activated and contaminated state. In this condition, the removal of the TFTR and its subsystems is greatly complicated. In 1999, PPPL commenced the TFTR Decontamination and Decommissioning (D&D) Project. The objectives of the D&D projects are to completely remove the TFTR in a safe, efficient and cost effective manner. To date all diagnostics, structural components and ancillary systems have been removed, and the tokamak is in the process of being sectioned into smaller pieces. This paper will describe the criteria, the methods employed and the successes of the D&D Project.

Recent experimental results in the fusion community have pointed to possible directions for improvement of the tokamak concept. A critical issue for many advanced tokamak scenarios is the optimization of the spatial distribution of plasma current density. Recent experiments on the Tokamak Fusion Test Reactor and Doublet III-D have demonstrated enhanced confinement and improved stability using transient techniques for the modification of the plasma current profile. The Princeton Beta Experiment-Modification (PBX-M) has begun experiments using profile modification techniques that are extrapolable to long-pulse or steady-state operation. PBX-M uses lower hybrid current drive for current profile modification, ion Bernstein wave heating for pressure profile variation, and plasma shaping to alter the magnetic geometry. This paper is an overview of the PBX-M program, its engineering facilities, and discussion of how this program is relevant to future advanced tokamak programs such as the Tokamak Physics Experiment.

The PBX-M passive plate stabilization system provides significant stabilization of long-wavelength external kink modes, the slowing of vertical instability growth rates, and the amelioration of disruption characteristics. The passive plate stabilization system has allowed the use of LHCD and IBW to induce current density and pressure profile modifications, and m=1 divertor biasing for modifying edge plasma transport. Improvements in the passive plate system insulators and support structures have provided reliable operation. Impurity influxes with the close-fitting passive plates are low. Solid target boronization is applied routinely to reduce conditioning time and maintain clean conditions.

PBX-M upgrades have been in progress for enhanced capability and reliability. These include upgrades to the vessel O-ring system for impurity control, the Ion Bernstein Wave and Lower Hybrid Current Drive Systems for plasma profile control, a double viewing upper and lower CHERS based Poloidal Rotation Diagnostic, and a new digital plasma control system.