C. Neumeyer

ITER power supply innovations and advances

The ITER Power Supply it will be the largest ever built in terms of power, pulse length, and energy capacity. It will also be responsible for fast discharge of the ITER superconducting magnets whose energy storage will be at an unprecedented scale. Nearly all of the components that comprise the system will be unique, custom designed items that will exceed the prior state of the art. This paper describes the ITER power supply system and its components with an emphasis on the extrapolation in scale and technology compared to the TFTR/JET/JT-60/T-15 era of large tokamaks with normal (copper) magnets as well as the present fleet of superconducting tokamaks (EAST, KSTAR). The main design issues, the chosen design solutions, and the collaborative design process will be described. The present state of development of the components by the Domestic Agencies is summarized, and areas of technological advancement are highlighted.

Design analysis and manufacturing studies for ITER In-Vessel Coils

ITER is incorporating two types of In Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required so that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are in close proximity to the plasma, mounted just behind the Blanket Shield Modules. This location results in a radiation and temperature environment that is severe necessitating new solutions for material selection as well as challenging analysis and design solutions. Fitting the coil systems in between the blanket shield modules and the vacuum vessel leads to difficult integration with diagnostic cabling and cooling water manifolds. The design of the IVCs is now progressing towards a final design scheduled for late CY 13. The project is a collaboration between the Princeton Plasma Physics Laboratory in Princeton NJ, the Chinese Academy of Sciences (ASIPP) in Hefei China and the ITER Organization. An extensive thermal and stress analysis to evaluate the effects of the high temperatures and electromagnetic loads on the In Vessel Coils has been undertaken. Manufacturing development is underway at ASIPP to develop the processes necessary to build ELM coil and VS Coil prototypes. This paper will outline the design and analysis issues as well as review the manufacturing development required to address these requirements and plans for prototypes.

NSTX TF joint failure and re-design

The toroidal field (TF) coil of the National Spherical Torus Experiment (NSTX) suffered a failure on February 14, 2003, after approximately 3 years of service and 7200 pulses. The failure occurred at an electrical joint connecting one of the radial flags to the inner leg bundle. Damage was extensive such that the coil could not be repaired. Analysis of the failure revealed structural design problems which have been addressed in the re-design effort described herein. Construction of a new TF coil is now underway.

Power Supply for NSTX Resistive Wall Mode Coils

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. Until 2004, the NSTX power system was feeding twelve (12) circuits in the machine. In 2004, resistive wall mode (RWM) coils were installed in the machine to suppress resistive wall modes and to correct error fields. There are six of these coils installed around the machine on the mid-plane. Since these coils need fast and accurate controls, a switching power amplifier (SPA) has been procured, installed and commissioned along with other circuit components. One of the existing thyristor rectifiers is used as the DC source to the SPA. The controls for the RWM have been integrated into the overall computer control of the DC power systems for NSTX. This paper describes the RWM power supply for NSTX

Quench protection circuits for superconducting magnets

In developing a scheme for the quench protection of the toroidal field (TF) and poloidal field (PF) superconducting magnets of the Tokamak Physics Experiment (TPX), an extensive review was performed of the design options and their performance characteristics. The general results and conclusions of these studies are reported herein. For tokamak magnets which have low enthalpy compared to their stored energy, quench protection requires the discharge of the stored energy at location(s) external to the magnets, typically in discharge resistors. Such discharge requires the interruption of large DC currents and the insertion of resistors using suitable DC circuit breaking methods. Since protection of the magnets is a crucial function, the system must be ultra-reliable, and new techniques are necessary.

Hardwired control system changes for NSTX DC power feeds

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The original TFTR Hardwired Control System (HCS) with electromechanical relays was used for NSTX DC Power loop control and protection during NSTX operations. As part of the NSTX Upgrade, the HCS is being changed to a PLC-based system with the same control logic. This paper gives a description of the changeover to the new PLC-based system.

Model for TFTR motor-generator (MG)

This study is aimed at predicting steady-state and dynamic responses after a sudden load shed for the TFTR motor-generator (MG) system. In the paper, a discussion on the methods, assumptions, and validation of the MG computer model is presented. The model includes the salient pole features (using two-axis theory) and effects of saturation. The steady-state model describing MG performance under normal conditions ignores the changes in flux linkage of windings other than the field winding. This simplification reduces the complexity of the model, yet it still describes the regular pulses of the TFTR generator satisfactorily. Only the field relation is described by the differential equation, and the rest are algebraic. The dynamic response of load shed-a special case in dynamic study-can be of importance in predicting the behavior of the MG system associated with the severe overvoltage problem. More elaborate synchronous generator models are required in this case. Not only the field winding voltage relation, but also the damper winding voltage relation must be described by differential equations. The complete solution can be obtained by means of the Laplace transform. Validation of the MG computer model has been performed by comparison with actual MG load scenarios recorded on electronic digitizers for TFTR shots. The simulation results are comparable to the recorded MG performance data.

Power supply changes for NSTX Resistive Wall Mode Coils

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. Prior to 2004, the NSTX power system was feeding twelve (12) circuits in the machine. In 2004 the Resistive Wall Mode (RWM) Coils were installed on the machine to correct error fields. There are six of these coils installed around the machine in the mid-plane. Since these coils need fast and accurate controls, a Switching Power Amplifier (SPA) with three sub-units was procured, installed and commissioned along with other power loop components. Two RWM Coils were connected in series and fed from one SPA sub-unit. After the initial RWM campaign, operational requirements evolved such that each of the RWM coils now requires separate power and control. Hence a second SPA with three sub-units has been procured and installed. The second unit is of improved design and has the controls and power components completely isolated. The existing thyristor rectifier is used as DC Link to both of the Switching Power Amplifiers. The controls for the RWM are integrated into the overall computer control of the DC Power systems for NSTX. This paper describes the design changes in the RWM Power system for NSTX.

Design description of the coaxial helicity injection (CHI) system on NSTX-U

Elimination of the central solenoid would simplify the engineering design of a Fusion Nuclear Science Facility (FNSF) and tokamak based devices. The method of Transient Coaxial Helicity Injection (CHI) has successfully demonstrated formation of a high-quality closed flux plasma in NSTX and will be used as the front end of the start-up method for a full demonstration of non-inductive current start-up, followed by non-inductive current ramp-up using neutral beams in the NSTX-U device that is now under construction at the Princeton Plasma Physics Laboratory (PPPL). CHI is implemented by driving current along open field lines that connect the lower inner and outer divertor plates of a Spherical Torus (ST). The engineering system requirements and the design of the CHI system on NSTX-U are described.

Digital Coil protection system for the National Spherical Torus experiment upgrade

An upgrade to the National Spherical Torus Experiment (NSTX) is currently in progress. The NSTX Center Stack Upgrade (NSTX-U) experimental device has an operating space that is both larger and more complex than that of the original NSTX. The mechanical integrity of some machine components can be compromised both by the instantaneous values of combinations of magnet currents and as a result of the time histories thereof. An upgrade to the existing protection systems and methodology is required to allow for both safe and effective use of the expanded operating space. A Digital Coil Protection System (DCPS) is planned as a major component of an upgraded Coil Protection System (CPS).