Carl J. Martin

THE ARIES-CS COMPACT STELLARATOR FUSION POWER PLANT

Abstract An integrated study of compact stellarator power plants, ARIES-CS, has been conducted to explore attractive compact stellarator configurations and to define key research and development (R&D) areas. The large size and mass predicted by earlier stellarator power plant studies had led to cost projections much higher than those of the advanced tokamak power plant. As such, the first major goal of the ARIES-CS research was to investigate if stellarator power plants can be made to be comparable in size to advanced tokamak variants while maintaining desirable stellarator properties. As stellarator fusion core components would have complex shapes and geometry, the second major goal of the ARIES-CS study was to understand and quantify, as much as possible, the impact of the complex shape and geometry of fusion core components. This paper focuses on the directions we pursued to optimize the compact stellarator as a fusion power plant, summarizes the major findings from the study, highlights the key design aspects and constraints associated with a compact stellarator, and identifies the major issues to help guide future R&D.

DESIGNING ARIES-CS COMPACT RADIAL BUILD AND NUCLEAR SYSTEM : NEUTRONICS, SHIELDING, AND ACTIVATION

Abstract Within the ARIES-CS project, design activities have focused on developing the first compact device that enhances the attractiveness of the stellarator as a power plant. The objectives of this paper are to review the nuclear elements that received considerable attention during the design process and provide a perspective on their successful integration into the final design. Among these elements are the radial build definition, the well-optimized in-vessel components that satisfy the ARIES top-level requirements, the carefully selected nuclear and engineering parameters to produce an economic optimum, the modeling - for the first time ever - of the highly complex stellarator geometry for the three-dimensional nuclear assessment, and the overarching safety and environmental constraints to deliver an attractive, reliable, and truly compact stellarator power plant.

SAFETY ASSESSMENT OF THE ARIES COMPACT STELLARATOR DESIGN

Abstract ARIES-CS is a 1000 MW(electric) compact stellarator conceptual fusion power plant design. This power plant design contains many innovative features to improve the physics, engineering, and safety performance of the stellarator concept. ARIES-CS utilizes a dual-cooled lead lithium blanket that employs low-activation ferritic steel as a structural material, with the first wall cooled by helium and the breeding zone self-cooled by flowing lead lithium. In this paper we examine the safety and environmental performance of ARIES-CS by reporting radiological inventories, decay heat, and radioactive waste management options and by examining the response of ARIES-CS to accident conditions. These accidents include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant event, and an in-vessel loss of coolant with bypass event that mobilizes in-vessel radioactive inventories (e.g., tritium and erosion dust from plasma-facing components). Our analyses demonstrate that the decay heat can be safely removed from ARIES-CS and the facility can meet the no-evacuation requirement.

Prediction of fabrication distortions in step and flash imprint lithography templates

Step and flash imprint lithography (SFIL) is an alternative approach for printing sub-100 nm features that relies on chemical and mechanical techniques to transfer patterns. The imprint process requires no projection optics and is performed at room temperature with low imprint pressures to reduce thermal and mechanical template distortions. Because imprint lithographies are 1× pattern transfer processes that preclude magnification corrections, the minimization of template distortions during fabrication and imprinting is critical. The processes and materials used in the fabrication of SFIL templates are similar to those used in the manufacturing of optical masks. The various process steps have been simulated using finite element techniques in order to predict the resulting pattern distortions. Two proposed template fabrication schemes were modeled, the resulting pattern distortions compared, and the error sources were quantified.

The ARIES Advanced and Conservative Tokamak Power Plant Study

Abstract Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5,a βtotal N of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m2. The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced-activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βtotal N of 2.5, an H98 of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.

ARIES-CS Magnet Conductor and Structure Evaluation

Abstract The ARIES-CS study focusing on the conceptual design and assessment of a compact stellarator power plant identified the important advantages and key issues associated with such a design. The coil configuration and structural support approach represent key design challenges, with the final design and material choices affected by a number of material and geometry constraints. This paper describes the design configuration and analysis and material choices for the ARIES-CS magnets and its structure. To meet aggressive cost and assembly/maintenance goals, the magnets are designed as lifetime components. Due to the very complex geometry, one of the goals of the study was to provide a robust operational design. This decision has significant implications on cost and manufacturing requirements. Concepts with both conventional and advanced superconductors have been explored. The coil structure design approach adopted is to wind all six modular coils of one field period in grooves in one monolithic coil structural shell (one per field period). The coil structural shells are then bolted together to form a strong structural shell to react the net radial forces. Extensive engineering analyses of the coil system have been performed using ANSYS shell and solid modeling. These include electromagnetic (EM) analyses to calculate the magnetic fields and EM forces and structural analyses to evaluate the structural responses and optimize the coil support system, which has a considerable impact on the cost of the ARIES-CS power plant.

Activation and Environmental Aspects of ARIES-ACT1 Power Plant

Abstract This paper reports the main results of the activation analysis of the ARIES-ACT1 power plant and its attractive environmental characteristics. The approach aims at examining several activation-related effects that influence the neutron-induced radioactivity, decay heat, waste classification, recyclability, and clearability of radioactive materials. Detailed studies are presented and provide information about the interdependence of material choices and activation response functions. It is also shown how the activation analysis can help address the inboard decay heat problem, one of the most important safety-related issues for tokamaks, and point the way to proper solutions through redefining the radial build to meet the design needs. Furthermore, the study suggests an integral management strategy to change what is now a costly waste disposal concern for fusion into a valued commodity via the further development of the recycling and clearance approaches.

Thermomechanical Analysis for an All-Tungsten ARIES Divertor

Abstract The ARIES team is currently proposing two tungsten divertor concepts for its tokamak designs and has performed extensive analyses to optimize their thermal and structural performance. Because of the high divertor operating temperatures and the low ductility of tungsten, thermal creep and fracture will be important failure mechanisms to consider. This paper presents a series of finite element analyses addressing the viable operating ranges of these tungsten plate divertor concepts with respect to creep and fracture. For fracture, the J-integral, a path-independent contour integral that estimates the strain energy release rate for a crack of assumed geometry, is used to address crack propagation. Elliptical surface cracks are introduced both inside and outside the coolant channel, and steady-state calculations are carried out for both full-power and cold shutdown conditions. It is determined that the critical crack is on the inside of the coolant channel with the highest stress intensities at full-power operation. Also, transient creep simulations are performed to predict the high-temperature thermal deformations and creep strains at various surface flux levels. Finally, transient thermal calculations are carried out to simulate edge-localized modes in the plasma, and conclusions are drawn with respect to the severity and frequency of these events with respect to surface melting for the two concepts.

Thermomechanical modeling of the EUV reticle during exposure

Thermal deformations of lithographic reticles during the exposure process may become an important consideration for all candidate Next-Generation Lithography technologies as these reticles are subject to stringent image placement and flatness requirements. The reflective reticles used for extreme ultraviolet lithography (EUVL) absorb energy during exposure producing temperature gradients and thermomechanical distortions that result in pattern placement errors. As throughput requirements of EUVL are increased, the necessary illumination power levels rise producing higher reticle temperatures. The use of a low-thermal-expansion substrate material reduces, but does not eliminate, reticle distortions, and the thermal and structural boundary conditions greatly influence the thermomechanical response. These factors make the accurate predictions of the reticle thermal and structural response essential to the design of EUVL systems. Previously published analyses focused on relatively low throughputs, 10 wafers-per-hour (wph), and 200-mm diameter wafer substrates. Proposed production systems have throughputs of 80 wph and use 6-in. square substrates. Finite element models of current format EUV reticles have been developed to simulate the reticles thermomechanical response to high-throughput exposure heating and assess the resulting image placement errors. The results of thermal and structural analyses for a variety of EUVL load and boundary conditions are presented.

Ratcheting Models for Fusion Component Design

Abstract One of the primary failure mechanisms addressed by structural design rules for fusion components is ratcheting, the accumulation of strain with cyclic loads. If a component is loaded such that ratcheting occurs, failure can be expected in relatively short order, so design rules must ensure that the behavior is avoided. In this paper, we present finite element models for cyclic loading of typical fusion structures and compare the results to analytical models for simple geometries and design rules intended for more complex geometries. Both material and structural ratcheting is considered. For structural ratcheting, the 3Sm rule employed in the ITER Structural Design Criteria is found to be unduly conservative and the accompanying Bree rules are found, in some cases, to be non-conservative. Significant advantage can be gained from using fully plastic models to avoid ratcheting.