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UWFDM-1084 Field-Reversed Configuration Power Plant Critical OLO HN GY C • Field-Reversed Configuration E I T N • S T Power Plant Critical Issues N I T O U I T S E U F € • W I N I S C O N S J.F. Santarius, G.A. Emmert, H.Y. Khater, E.A. Mogahed, C.N. Nguyen, L.C. Steinhauer, G.H. Miley June 1998 UWFDM-1084 FUSION TECHNOLOGY INSTITUTE UNIVERSITY OF WISCONSIN MADISON WISCONSIN DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government, nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. FIELD-REVERSED CONFIGURATION POWER PLANT CRITICAL ISSUES J.F. Santarius, G.A. Emmert, L.C. Steinhauer G.H. Miley H.Y. Khater, E.A. Mogahed, and Redmond Plasma Physics Lab Fusion Studies Lab C.N. Nguyen University of Washington University of Illinois Fusion Technology Institute 14700 NE 95th St., Suite 100 100 NEL University of Wisconsin Redmond, WA 98052 103 S. Goodwin Ave. 1500 Engineering Dr. (425)881-9380 Urbana, IL 61801-2984 Madison, WI 53706 (217)333-3772 (608)263-2252 ABSTRACT shield design, economics, plasma particle and power balance, current drive, plasma-surface interactions, and systems integration. Both D-T and D-3He fuels An Engineering Scoping Study of a Field-Reversed appear likely to perform well in FRC power plants, Configuration (FRC) burning D-T fuel is being but the focus of the present study is on D-T fuel. performed by the Universities of Wisconsin, Washington, and Illinois. The effort concentrates on With regard to fusion development, the tritium-breeding blanket design, shielding, radiation FRC provides a good balance between physics damage, activation, safety, environment, plasma uncertainty and engineering attractiveness. The modeling, current drive, plasma-surface interactions, trade-offs among physics, engineering, safety, and economics, and systems integration. A systems environmental considerations have only recently analysis code will serve as the key tool in defining gained prominence|partly due to the difficulties a reference point for detailed physics and engineering encountered when the fusion community realistically calculations plus parametric variations. Advantages faced engineering issues in designing ITER. While of the cylindrical geometry and high β (plasma the physics obstacles on the FRC development path pressure/magnetic-field pressure) are evident. should not be underestimated, excellent progress is being made by the small worldwide FRC research community.1 The key physics issues include operation I. INTRODUCTION at large s (average number of radial gyroradii), startup with reasonable power, and sustainment. From an engineering standpoint, an FRC burning Field-reversed configuration (FRC) power plants D-T fuel appears capable of being built with appear likely to provide an excellent balance between near-term technology to a large extent. The potential reactor attractiveness and technical main exceptions are the materials used for the development risk. In particular, (1) the linear, first wall, blanket, and shield, which will be cylindrical FRC geometry facilitates the design of subject to high neutron fluences with consequent tritium-breeding blankets, shields, magnets, and radiation damage and activation. If the more input-power systems, and (2) the high FRC β difficult physics requirements of D-3He fuel can be (plasma pressure/magnetic field pressure) increases achieved, essentially all necessary FRC technology the plasma power density and allows a compact appears to be in hand, benefits would be gained reactor design. The surface heat flux is moderate, from direct conversion, and environmental and safety however, because much of the fusion power is carried characteristics would be substantially improved.2 by the plasma flowing to the end chamber walls. The present research project investigates critical Excellent recent progress in FRC physics issues for FRC fusion power plants, concentrating motivates the present work and makes the mainly on engineering areas. The issues being research especially timely. Highlights include studied include tritium-breeding blanket design, indications that natural minimum-energy FRC states radiation damage, activation, safety, environment, exist, demonstration of stability to global MHD 1 modes, stable operation at moderate s (plasma has been modified to include FRC physics, radius/average gyroradius), startup by merging two engineering, and economics based on University spheromaks to form an FRC, and current drive by of Washington physics models and University of rotating magnetic fields.1 Wisconsin engineering and economics models. The University of Wisconsin is developing an • innovative tritium breeding blanket concept II. OBJECTIVE AND TASKS and evaluating the possible use of previously published blanket designs. The objective of the present scoping study is to The University of Wisconsin is assessing investigate the critical engineering issues for D-T • options for radiation shielding and evaluating FRC electric power plants. The main tasks involved required shield thicknesses. in this research and the institutions with primary responsibility are: The University of Washington has generated • plasma-physics and current-drive models. These University of Wisconsin will be refined as the study progresses. • The University of Illinois is investigating 2 Coordination • plasma-surface interactions and plasma exhaust handling. 2 Systems analysis and economics 2 Tritium-breeding blanket design A. Plasma and Current Drive Modeling 2 Radiation shielding and damage The initial calculations use relatively simple 2 Activation, safety, and environment quasi-equilibrium profiles: 1 University of Washington tanh− σu • U(u) κ (1) ≡ σ 2 Plasma modeling r2 u(r; z) 2 1(2) 2 Current drive ≡ R − 1 4 4 2 R(z) R0 1 z =b ) (3) University of Illinois ≡ − • B(r; z)=Be tanh U (4) 2 Plasma-surface interactions cBe 2r dU 2 jθ(r; z)= sech U (5) 4π · R2(z) du 2 Plasma exhaust handling 2 p(r; z)=pmsech U (6) 2=Γ n(r; z)=nm(sech U) (7) If time and resources permit, other areas will be 2(Γ 1)=Γ investigated. Those with the potential for having T (r; z)=Tm(sech U) − (8) an important positive or negative impact on the where r; θ; z define a cylindrical coordinate system, design include maintenance, high-Tc superconducting 1 f g a =22 R = separatrix radius, b = FRC half length, magnets, energy conversion, liquid-metal first walls, 0 R = radius of the 0-point at the midplane, j = and systems integration. 0 current density, p = pressure, n =density,T = temperature, c = speed of light, Be =external magnetic field, and the subscript m signifies the III. STATUS maximum value. κ and σ are small shaping parameters, and Γ 4=3. ∼ The status of the study as of June, 1998, which is about nine months into the two-year project, is as Viable FRC startup and sustainment methods follows: with reasonable input powers are being sought. The requirements for two promising methods, rotating- The University of Wisconsin systems code, magnetic-field (RMF) current drive and merging • which was written initially for tokamaks, spheromaks, are being determined and modeled. 2 High-Temperature He Exit Superconducting Manifold He/Water Magnet A Biological Shield Segmented Steel/Water Shield He Supply Segmented Blanket (Li2O/SS [HT-9]) Target Plate He Exit A @ 5 m He Supply Section A-A One-half View Figure 1: One-half view and cross section of a preliminary concept for the tritium-breeding blanket. B. Tritium-Breeding Blanket Design coolant would be used for the shield. Another possibility is use of the Mirror Advanced Reactor 3 A tritium-breeding blanket for an FRC is in many Study (MARS) blanket, shown in Fig. 2, with key ways simpler than for a tokamak. The extremely parameters given in Table 1. This approach would high tokamak magnetic fields lead to large toroidal take advantage of the geometrical similarity of FRC field coils which, along with the toroidal geometry, and tandem-mirror core regions. reduce maintenance access and usually require splitting blanket modules into several submodules and translating them toroidally for removal. In an FRC, the cylindrical geometry and low magnetic field allow removal of single modules containing the first wall, blanket, shield, and magnets. If liquid-metal coolants are used, the MHD pressure drop will also be substantially reduced by the low magnetic field and short flow paths. A key question is whether it will be necessary to use materials requiring long-term development, such as SiC or V, or use can be made of nearly off-the-shelf materials, such as low-activation ferritic and austenitic steels. We are pursuing parallel courses for scoping tritium-breeding blanket designs: (1) evaluate new ideas and (2) assess established concepts. Figure 1 shows the initial blanket concept, which uses Li2O breeder, austenitic or ferritic steel structure, and Figure 2: Cross section of the MARS tritium-breeding helium coolant. Stainless steel structure and water blanket. 3 Table 1: Parameters for the MARS Blanket. Table 2: MINIMARS central cell parameters. Structure HT-9 Steel Number of modules 24 Coolant Li17 Pb83 Module length 2.8 m Breeder Li17 Pb83 Module radius 1.75 m rw 0.6 m Magnet thickness 0.06 m 2 Γn 4.3 MW/m Peak B field 3.1 T M 1.36 Current density 37 MA/m2 TBR 1.15 Current 30 kA ηth 42% the steady-state and much higher peak and average C.
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