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This paper summarizes the results of a detailed study reported in Ref. [1]. A concept for a 1000 MW(e) Z-Pinch IFE power plant has ten reactor chambers. A Load, consisting of a wire array surrounding a cryogenic dynamic hohlraum target with a 3-GJ yield, must be produced, inserted into a Replaceable Transmission Lines (RTL), transported to one of the chambers, and shot every second. The conceptual design of the load facilitates automatic mass manufacturing and insertion at 1 Hz at minimum cost. A sequence of operations necessary to (1) fabricate each wire array on a holding and insertion tool, (2) manufacture, evacuate, fill with DT, freeze, layer, and assemble each target with helium gas and liquid hydrogen, (3) insert a wire array and a target under vacuum into an RTL, and (4) remove their debris, is described. Detailed cost estimates derived for the complete load production and solid debris removal cycle in a commercial-scale facility are summarized.

AbstractThe research and development of manufacture-related technology for ITER magnet supports is one of the tasks for construction. AISI 316LN austenitic stainless steel has been developed and tested as the main raw material. The material shows excellent mechanical properties at room temperature, 77 K, and 4.2 K. An alternative design for the toroidal field support manufacture without welding was carried out. The structural analysis shows no stress concentration and buckling in the present design during ITER operation. However, further engineering tests of the structural stability under various load combinations are also scheduled. A brazed connection to attach the cooling pipes to the support plates is suggested. Several kinds of candidate brazing fillers, such as Sn-Pb-, Ag-, and Cu-based alloys have been developed. The tensile strength of the brazed solders is up to 400 MPa at 77 K for the Ag-based and Cu-based fillers. For correction coil support, the plasma spray insulation coating was developed and introduced.

AbstractThe Toroidal Field (TF) system of ITER consists of 18 coils with design nominal current of 68 kA operating in steady state mode that provides 5.4 T in the plasma centre. The winding pack (WP) of each coil is formed by 7 stacked double pancakes which are connected between them in the coil lower region, sharing space with the current leads, supercritical Helium cooling piping and manifolds. The TF coils of ITER are not nuclear safety related, but the release of the 41 GJ of magnetic energy in a controlled way in case of a quench and the difficulties of replacing a failing TF coil make a reliable coil instrumentation design essential as investment protection.The present paper describes not only the principles of the primary and secondary quench detection system of the ITER TF coils but also the operation monitoring instrumentation. The reliability of strain gauges, temperature sensors, pressure gauges and flow meters in the cryogenic environment and high electromagnetic noise environment is also discussed.

AbstractIn a central building of the power plant, the DT fuel is formed into a very smooth and uniform layer of ice at ˜18 K inside a beryllium shell; placed in a cryogenic target assembly that provides support, cooling, and thermal insulation; and put into an evacuated replaceable transfer line (RTL) at room temperature (RT). The RTL is transported and inserted into one of the reactor chambers at 923 K and shot, releasing 3 GJ of nuclear fusion energy. The DT ice layer must stay below ˜19.7 K to keep its geometric integrity until shot time.Detailed transient thermal analyses of the cryogenic target assembly in the RTL were performed. They showed that, with the original design, the DT ice would reach 24.6 K by shot time. With an improved design providing better thermal insulation of the target, the ice temperature would reach only 19.1 K, meeting the requirement for successful shots.This paper compares the thermal analysis results for both designs, which included conduction and radiation effects with temperature-dependent material properties.

AbstractFor electricity production in a 1000 MW(e) Z-Pinch Inertial Fusion Energy (IFE) power plant, a wire array must be produced and shot every second. The slow and painstaking manual assembly and insertion process developed for the present Z-Pinch experimental machines will have to be replaced with mass production and rapid auto matic handling. This could be facilitated by making one-piece, or unitized, wire arrays (UWA). This paper reviews potential UWA manufacturing processes; describes the results of etching, milling, laser-cutting, and lithography tests applied to an UWA design that could be shot on the Z-R machine for validation; assesses the feasibility of these processes for mass production; and proposes an alternate UWA design concept for easier manufacturing.

AbstractThe hohlraum surrounds the fuel capsule in a heavy ion fusion (HIF) target. The hohlraum absorbs ion beam driver energy and emits this energy uniformly around the capsule in the form of X-rays. High-atomic-number materials are necessary in the walls of the hohlraum to contain the X-ray energy around the capsule during the implosion process. These high-atomic-number hohlraum materials affect many aspects of an HIF power plant operation. A systematic review of available information for all high-atomic-number elements was conducted to select candidate hohlraum materials. The effects of materials on target fabrication, energy cost, target gain, radioactivity, chemical toxicity, and potential for recycle were considered. Lead and tungsten are the lowest-cost acceptable materials in the primary coolant. The combination of lead and tungsten provide better target gain than either material alone. Seeding the primary coolant with submicron-sized tungsten particles can minimize tungsten growth in small openings in power plant components such as vacuum tritium disengagers. Concerns remain for possible tungsten particle agglomeration, settling, or erosion caused by tungsten particles. Tungsten could be replaced by several lanthanide elements if tungsten proves unacceptable.

AbstractA central feature of an Inertial Fusion Energy (IFE) power plant is a target that has been compressed and heated to fusion conditions by the energy input of the driver. The technology to economically manufacture and then position cryogenic targets at chamber center is at the heart of future IFE power plants. For direct drive IFE (laser fusion), energy is applied directly to the surface of a spherical CH polymer capsule containing the deuterium-tritium (DT) fusion fuel at approximately 18K. For indirect drive (heavy ion fusion, HIF), the target consists of a similar fuel capsule within a cylindrical metal container or ’’hohlraum’’ which converts the incident driver energy into x-rays to implode the capsule. For either target, it must be accurately delivered to the target chamber center at a rate of about 5-10Hz, with a precisely predicted target location. Future successful fabrication and injection systems must operate at the low cost required for energy production (about $0.25/target, about 104less than current costs).Z-pinch driven IFE (ZFE) utilizes high current pulses to compress plasma to produce x-rays that indirectly heat a fusion capsule. ZFE target technologies utilize a repetition rate of about 0.1 Hz with a higher yield.This paper provides an overview of the proposed target methodologies for laser fusion, HIF, and ZFE, and summarizes advances in the unique materials science and technology development programs.

AbstractThe long-range goal of the Z-Pinch IFE program is to produce an economically-attractive power plant using high-yield z-pinch-driven targets (~3GJ) with low rep-rate per chamber (~0.1 Hz). The present mainline choice for a Z-Pinch IFE power plant uses an LTD (Linear Transformer Driver) repetitive pulsed power driver, a Recyclable Transmission Line (RTL), a dynamic hohlraum z-pinch-driven target, and a thick-liquid wall chamber. The RTL connects the pulsed power driver directly to the z-pinch-driven target, and is made from frozen coolant or a material that is easily separable from the coolant (such as carbon steel). The RTL is destroyed by the fusion explosion, but the RTL materials are recycled, and a new RTL is inserted on each shot.A development path for Z-Pinch IFE has been created that complements and leverages the NNSA DP ICF program. Funding by a U.S. Congressional initiative of $4M for FY04 through NNSA DP is supporting assessment and initial research on (1) RTLs, (2) repetitive pulsed power drivers, (3) shock mitigation [because of the high yield targets], (4) planning for a proof-of-principle full RTL cycle demonstration [with a 1 MA, 1 MV, 100 ns, 0.1 Hz driver], (5) IFE target studies for multi-GJ yield targets, and (6) z-pinch IFE power plant engineering and technology development. Initial results from all areas of this research are discussed.

In inertial fusion energy (IFE) power plant designs, the fuel is a spherical layer of frozen DT contained in a target that is injected at high velocity into the reaction chamber. For direct drive, typically laser beams converge at the centre of the chamber (CC) to compress and heat the target to fusion conditions. To obtain the maximum energy yield from the fusion reaction, the frozen DT layer must be at about 18.5 K and the target must maintain a high degree of spherical symmetry and surface smoothness when it reaches the CC. During its transit in the chamber the cryogenic target is heated by radiation from the hot chamber wall. The target is also heated by convection as it passes through the rarefied fill-gas used to control chamber wall damage by x-rays and debris from the target explosion. This article addresses the temperature limits at the target surface beyond which target uniformity may be damaged. It concentrates on direct drive targets because fuel warm up during injection is not currently thought to be an issue for present indirect drive designs and chamber concepts. Detailed results of parametric radiative and convective heating calculations are presented for direct-drive targets during injection into a dry-wall reaction chamber. The baseline approach to target survival utilizes highly reflective targets along with a substantially lower chamber wall temperature and fill-gas pressure than previously assumed. Recently developed high-Z material coatings with high heat reflectivity are discussed and characterized. The article also presents alternate target protection methods that could be developed if targets with inherent survival features cannot be obtained within a reasonable time span.

The ITER magnet system consists of structurally linked sets of toroidal (TF) and poloidal (PF) field coils, central solenoid (CS), and various support structures. The coils are superconducting, force flow Helium cooled with a Kapton-Glass-Epoxy multilayer insulation system. The stored magnetic energy is about 100GJ in the TF system and 20GJ in the PF-CS. Coils and structure are maintained at 4 K by enclosing them in a vacuum cryostat. The cryostat, comprising an outer envelope to the magnets, forms most of the second radioactivity confinement barrier. The inner primary barrier is formed by the vacuum vessel, its ports and their extensions. To keep the machine size within acceptable bounds, it is essential that the magnets are in close proximity to both of the nuclear confinement barriers. The objective of the magnet design is that, although local damage to one of the barriers may occur in very exceptional circumstances, large scale magnet structural or thermal failure leading to simultaneous breaching of both barriers is not credible. Magnet accidents fall into three categories: thermal (which includes arcing arising from insulation failure and local overheating due to discharge failure in the event of a superconductor quench), structural (which includes component mechanical failure arising from material inadequacies, design errors and exceptional force patterns arising from coil shorts or control failures), and fluid (Helium release due to cooling line failure). After a preliminary survey to select initial faults conceivable within the present design, these faults are systematically analyzed to provide an assessment of the damage potential. The results of this damage assessment together with an assessment of the reliability of the monitoring and protective systems, shows that the magnets can operate with the required safety condition.

ITER will be the first large-scale tokamak to be designed as a nuclear facility to provide public protection from external hazards such as earthquakes. The design approach for such events has been developed consistent with ITER's moderate hazards and overall safety approach on a basis of the ITER site assumptions. Seismic design is described including selection of ground motions for design purposes, seismic safety requirements, and the seismic classification scheme. The results of preliminary seismic assessments are summarized including the potential for seismically induced plasma vertical displacement events (VDE). Finally, potential facility modifications available to deal with site-specific external hazards are suggested. At the Detailed Design Report stage of the Engineering Design Activity (EDA), it is concluded that ITER has been designed to deal with the site design assumptions for earthquakes and can be designed to safety cope with a range of site-specific external hazards with modest changes to the facility.

The ITER superconducting magnet system (Toroidal Field coils (TF), Central Solenoid (CS), Poloidal Field coils (PF), Correction Coils (CC) and Feeders) represents a total weight of approximately 10 000 tons. It is designed to withstand gravitational and seismic forces, stresses induced by thermal contractions during cool-down from 300 K to 4.5 K, and large Lorentz forces in the superconducting coils. Although not part of the Safety Important Class (SIC) components, the sensors (stress, displacement, thermometer) used to monitor the thermo-mechanical behavior of the structures are key diagnostic means to assess the design, support operation and survey possible fatigue effects over the 20 years' lifetime of the Tokamak. The sensors and their wiring will be submitted to unique and severe environmental operating conditions; they will have to operate in the cryostat's high vacuum (<;10-4 Pa total gas pressure) at magnetic inductions of several Tesla and at low temperature (4.5 K) in the presence of a neutron fluence on the order of 1022 m-2 -- which implies a gamma dose higher than 10 MGy -- for a period of 20 years. In addition, the fast and slow cycling of very large currents in the superconducting magnets and plasma will generate large eddy currents and, consequently, heat loads and electromotive forces (voltages) in the various parts of the structures, as well as a large electromagnetic noise. This paper presents the design philosophy and the chosen sensor technologies, some of which require further development in collaboration with the industry so that they can meet the objectives and the severe ITER operating conditions. In particular, only 20 percent of the near 1 000 measuring points for the thermo-mechanical data of the ITER magnet structures will rely on "classical" copper-wire technologies whereas 80 percent will rely on specifically developed optical-fiber-based sensors. [ABSTRACT FROM AUTHOR]