[en] The paper presents status of these development tasks, including winding development, inlets and outlets development, internal and bus joints development and testing, insulation development and qualification, vacuum-pressure impregnation, breakout regions, bus supports, TF conductor fabrication development, intermodule structure and materials characterization. The U.S. ITER Project Office (USIPO) is responsible for supplying the central solenoid (CS) with preload and support structure and nine lengths of the toroidal field (TF) conductor for the ITER machine currently under construction in Cadarache, France. Several features of the CS design needed to be developed and qualified during the development and testing activity prior to design completion and beginning fabrication. This is necessary because the CS is a unique and challenging solenoid, a significant step forward from the past achievements. Many aspects of the CS design are unprecedented in industry experience. The tolerances on the CS turn location are very tight, especially in the joggles region; therefore, the winding machine and auxiliary tools need to be developed to ensure feasibility of the design. The helium inlets are located in the area with the highest stress and magnetic field and the lowest temperature margin; therefore, they represent a significant fabrication and performance risk. The CS insulation needs to withstand up to 30 kV and remain structurally robust, which is a very challenging task that has never been addressed for fusion magnets in the past. The total weight of the CS assembly is about 1000 t; it is about 16m high and 4.3m in diameter. A detailed description of the CS design is given in Ref. (1). This paper discusses the most critical R and D tasks being managed by the USIPO. The ITER CS consists of 6 Modules stacked together with the CS structure that keeps it together under preload in the center of the ITER machine.

[en] U.S.-ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). TCWS is designed to provide cooling and baking for client systems that include the first wall/blanket, vacuum vessel, divertor, and neutral beam injector. Additional operations that support these primary functions include chemical control of water provided to client systems, draining and drying for maintenance, and leak detection/localization. TCWS interfaces with 27 systems including the secondary cooling system, which rejects this heat to the environment. TCWS transfers heat generated in the Tokamak during nominal pulsed operation - 850 MW at up to 150 C and 4.2 MPa water pressure. Impurities are diffused from in-vessel components and the vacuum vessel by water baking at 200-240 C at up to 4.4 MPa. TCWS is complex because it serves vital functions for four primary clients whose performance is critical to ITER's success and interfaces with more than 20 additional ITER systems. Conceptual design of this one-of-a-kind cooling system has been completed; however, several issues remain that must be resolved before moving to the next stage of the design process. The 2004 baseline design indicated cooling loops that have no fault tolerance for component failures. During plasma operation, each cooling loop relies on a single pump, a single pressurizer, and one heat exchanger. Consequently, failure of any of these would render TCWS inoperable, resulting in plasma shutdown. The application of reliability, availability, maintainability, and inspectability (RAMI) tools during the different stages of TCWS design is crucial for optimization purposes and for maintaining compliance with project requirements. RAMI analysis will indicate appropriate equipment redundancy that provides graceful degradation in the event of an equipment failure. This analysis helps demonstrate that using proven, commercially available equipment is better than using custom-designed equipment with no field experience and lowers specific costs while providing higher reliability. This paper presents a brief description of the TCWS conceptual design and the application of RAMI tools to optimize the design at different stages during the project.

[en] In its simplest form, risk management is a continuous assessment from project start to completion that identifies what can impact your project (i.e., what the risks are)., which of these risks are important, and identification and implementation of strategies to deal with these risks (both threats and opportunities). The National Compact Stellerator Experiment (NCSX) Project was a 'first-of-a-kind' fusion experiment that was technically very challenging, primarily resulting from the complex component geometries and tight tolerances. Initial risk quantification approaches proved inadequate and contributed to the escalation of costs as the design evolved and construction started. After the Project was well into construction, a new risk management plan was adopted. This plan was based on successful Department of Energy (DOE) and industrial risk management precepts. This paper will address the importance of effective risk management processes and lessons learned. It is of note that a steady reduction of risk was observed in the last six months of the project

[en] The paper presents the status of research and development (R and D) magnet tasks that are being performed in support of the U.S. ITER Project Office (USIPO) commitment to provide a central solenoid assembly and toroidal field conductor for the ITER machine to be constructed in Cadarache, France. The following development tasks are presented: winding development, inlets and outlets development, internal and bus joints development and testing, insulation development and qualification, vacuum-pressure impregnation, bus supports, and intermodule structure and materials characterization.

[en] The Columbia Nonneutral Torus is a new stellarator experiment being built at Columbia University, New York, to study the confinement of nonneutral and electron-positron plasmas. It will be a two-period, ultralow aspect ratio classical stellarator configuration created from four circular coils. The theory of the confinement and transport of pure electron plasmas on magnetic surfaces is reviewed. The guiding principles behind the experimental design are presented, together with the actual experimental design configuration