Brad J. Merrill

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.

ENGINEERING DESIGN AND ANALYSIS OF THE ARIES-CS POWER PLANT

Abstract The ARIES-CS team has concluded an integrated study of a compact stellarator power plant, involving physics and engineering design optimization. Key engineering considerations include the size of the power core, access for maintenance, and the minimum distance required between the plasma and the coil to provide acceptable shielding and breeding. Our preferred power core option in a three-field-period configuration is a dual-coolant (He + Pb-17Li) ferritic steel modular blanket concept coupled with a Brayton power cycle and a port-based maintenance scheme. In parallel with a physics effort to help determine the location and peak heat load to the divertor, we developed a helium-cooled W alloy/ferritic steel divertor design able to accommodate 10 MW/m2. We also developed an intercoil structure design to accommodate the electromagnetic forces within each field period while allowing for penetrations required for maintenance, plasma control, coolant lines, and supporting legs for the in-vessel components. This paper summarizes the key engineering outcomes from the study. The engineering design of the fusion power core components (including the blanket and divertor) are described and key results from the supporting analyses presented, including stress analyses of the components and thermal-hydraulic analyses of the power core coupled to a Brayton cycle. The preferred port-based maintenance scheme is briefly described and the integration of the power core is discussed. The key stellarator-specific challenges affecting the design are highlighted, including the impact of the minimum plasma-coil distance, the maintenance, integration, and coil design requirements, and the need for alpha power accommodation.

Modifications to the MELCOR code for application in fusion accident analyses

For the past several years, the Fusion Safety Program at the Idaho National Engineering and Environmental Laboratory (INEEL) has modified the MELCOR code in order to assess safety issues associated with loss-of-cooling accidents (LOCAs) and loss-of-vacuum accidents (LOVAs) in the international thermonuclear experimental reactor (ITER) engineering design activity (EDA). MELCOR is a thermal hydraulics computer code developed by the Sandia National Laboratory for analyzing severe accidents in fission power plants. This paper describes these modifications and the role they played in LOVA and LOCA analyses performed for the non-site specific safety report (NSSR) for the ITER EDA.

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.

The Fusion Nuclear Science Facility, the Critical Step in the Pathway to Fusion Energy

The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs.

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.

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

Abstract Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W–5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. Systems study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kW h. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

Modeling of particulate production in the SIRENS plasma disruption simulator

Abstract Modeling of the complex interplay among plasma physics, fluid mechanics, and aerosol dynamics is critical to providing a detailed understanding of the mechanisms responsible for particulate production from plasma–surface interaction in fusion devices. Plasma/fluid and aerosol models developed for analysis of disruption simulation experiments in the SIRENS high heat flux facility integrate the necessary mechanisms of plasma–material interaction, plasma and fluid flow, and particulate generation and transport. The model successfully predicts the size distribution of primary particulate generated in SIRENS disruption-induced material mobilization experiments.

Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

Abstract As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li2BeF4 and the low melting point molten salts such as LiBeF3 and LiNaBeF4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiCf/SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R&D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan.

A lithium-air reaction model for the melcor code for analyzing lithium fires in fusion reactors

Abstract This paper details the initial modifications made to the melcor code that allow it to predict the consequences of lithium spill accidents for evolving fusion reactor designs. These modifications include thermodynamic and transport properties of lithium, and physical models for predicting the rate of reaction of, and energy production from the lithium-air reaction. A benchmarking study was performed with this new melcor capability. Two lithium-air reaction tests conducted at the Hanford Engineering Development Laboratory (HEDL) were selected for this benchmark study. Excellent agreement was achieved between melcor predictions and the measured data, when several empirically derived coefficients were used. Recommendations for modeling lithium fires with melcor and for future work in this area concluded this paper.