Final Report for the Field-Reversed Configuration Power Plant Critical-Issue Scoping Study

Summary

This final report describes a scoping study of critical issues for field-reversed configuration (FRC) fusion power plants, focusing on deuterium-tritium (D-T) systems analysis, fusion core conceptual design, and engineering feasibility. It details the thermo-mechanical and thermal-hydraulic design of a helium-cooled solid breeder (Li2O) blanket and shield module using advanced ferritic steel, provides the theoretical foundation for the Wisconsin Systems Code (WISC), and includes comparative evaluations including thick liquid-wall configurations.

Cover Page

J.F. Santarius, E.A. Mogahed, G.A. Emmert, H.Y. Khater, C.N. Nguyen, S.V. Ryzhkov, M.D. Stubna, L.C. Steinhauer, G.H. Miley

March 2000

UWFDM-1129

FUSION TECHNOLOGY INSTITUTE UNIVERSITY OF WISCONSIN MADISON WISCONSIN

Disclaimer

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.

Title Page & Project Info

Final Report for the Field-Reversed Configuration Power Plant Critical-Issue Scoping Study

Project DE-FG02-97ER54431 September 1, 1997—December 31, 1999

John F. Santarius, Elsayed A. Mogahed, Gilbert A. Emmert, Hesham Y. Khater, Canh N. Nguyen, Sergei V. Ryzhkov, and Michael D. Stubna Fusion Technology Institute, Department of Engineering Physics, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, WI 53706

Loren C. Steinhauer University of Washington

George H. Miley University of Illinois

March 2000 UWFDM-1129

Table of Contents

List of Figures … iii List of Tables … iv

1 Introduction … 1-1 1.1 Project activities … 1-1 1.1.1 Systems analysis and fusion core conceptual design … 1-1 1.1.2 Talks and poster papers … 1-2 1.1.3 Papers … 1-3 1.1.4 Leveraged activities … 1-4 1.1.5 Student involvement … 1-5 1.1.6 Recommendations for future FRC power plant research … 1-6 1.2 Overview … 1-7 1.3 References for Section 1 … 1-8

2 Fusion Core Engineering … 2-1 2.1 The thermo-mechanical design of the first wall and blanket of the FRC … 2-1 2.2 The mechanical design of the first wall and blanket of the FRC … 2-3 2.3 Coolant routing … 2-6 2.4 Thermal hydraulics calculations … 2-7 2.5 Results of thermal hydraulics calculations of the FRC components … 2-9 2.6 Limits on the Li2O tube dimensions … 2-16 2.7 Maintenance … 2-17 2.8 References for Section 2 … 2-18

3 WISC: the WIsconsin Systems Code … 3-1 3.1 Introduction … 3-1 3.2 Nomenclature … 3-2 3.2.1 Subscripts … 3-2 3.2.2 Variables … 3-2 3.3 Description of the WISC code … 3-6 3.3.1 Initialization … 3-6 3.3.2 Charge … 3-7 3.3.3 Coulomb logarithm … 3-8 3.3.4 Geometric and density parameters … 3-9 3.3.5 Slowing down … 3-11 3.3.6 Critical energy … 3-11 3.3.7 Fast energy fraction going to ions … 3-12 3.3.8 Fusion reaction rate … 3-13 3.3.9 Fast ion pressure … 3-13 3.3.10 Energy confinement time … 3-14 3.3.11 Particle confinement time … 3-14 3.3.12 Hot-ion tail density … 3-15 3.3.13 Charged particle fusion power … 3-15 3.3.14 Fusion power going to electrons … 3-16 3.3.15 Bremsstrahlung power … 3-16 3.3.16 Rotating magnetic field current drive power … 3-17 3.3.17 Synchrotron power … 3-17 3.3.18 Transport power … 3-17 3.3.19 Electron temperature … 3-18 3.3.20 Density … 3-18 3.3.21 Edge plasma … 3-19 3.3.22 Blanket … 3-20 3.3.23 Shield … 3-20 3.3.24 Magnets … 3-20 3.3.25 Total radial build … 3-20 3.3.26 Net power … 3-21 3.3.27 Miscellaneous power … 3-22 3.4 References for Section 3 … 3-23

4 Reference Case and APEX Liquid-Wall Cases … 4-1 4.1 Reference case … 4-1 4.2 APEX case … 4-3 4.3 Reference for section 4 … 4-5

5 Appendix: Papers related to this work … 5-6

Section 1: Introduction

1 Introduction

1.1 Project activities This report describes research in which a team from the Universities of Wisconsin, Washington, and Illinois performed a scoping study of critical issues for field-reversed configuration (FRC) power plants. The key tasks for this research were:

Systems analysis of deuterium-tritium (D-T) FRC fusion power plants
Conceptual design of the blanket and shield module for an FRC fusion core. The effort of approximately one-third of a full time equivalent (FTE) professional researcher per year was split among the participating institutions as follows: University of Wisconsin (70%), University of Washington (20%), and University of Illinois (10%).

1.1.1 Systems analysis and fusion core conceptual design In order to perform the systems analysis, the University of Wisconsin’s fusion power plant systems code, described in detail in Section 3, was modified from its initial form as a tokamak systems code to include FRC physics and engineering models. Some alternate modes of operation were also explored. The reference case and a case generated assuming liquid walls and very high power density are given in Section 4. A key thrust of the research was to investigate a crucial question for FRC power plants: Given success in the physics, would the engineering features of the resulting device be attractive? For the engineering conceptual design of the FRC fusion core, therefore, the project team focused on intermediate-term technology. For example, one decision was to use steel structure instead of exotic but relatively undeveloped materials, such as vanadium or silicon carbide. The FRC does indeed appear to lead to an attractive fusion power plant, based on several features of the design, including modest size, cylindrical symmetry, good thermal efficiency, and relatively easy maintenance. The resulting compact FRC fusion core of the reference case conceptual design possesses a high ratio of electric power to fusion core mass, indicating that it would certainly have favorable economics. Details of the design are discussed in Section 2.

1.1.2 Talks and poster papers Table 1-1 lists the talks that project team members have given that are either directly or indirectly related to this research. Similarly, Table 1-2 lists related poster papers.

1.1.6 Recommendations for future FRC power plant research

The cylindrically symmetric geometry of FRC fusion cores allows the design of an attractive first-wall/blanket/shield/magnet module with reasonable engineering assumptions. The concept presented in this report should be pursued further for the FRC and other amenable configurations, such as the spheromak and spherical torus.
Field-reversed configurations show promise for providing the most attractive performance of any magnetic fusion concept, and a detailed, integrated, conceptual design of a D-T FRC power plant should be undertaken at the level of at least 10 FTE’s per year for two years.
The geometry of FRC fusion cores fits the use of liquid walls very well, and the combination should be investigated in more depth.
The FRC appears well suited to burning D-3He fuel, and a detailed conceptual design at several FTE’s per year for two years should be performed in order to assess whether such a device could achieve the watershed level of sufficiently attractive economics to break into the electricity market.
Other applications of FRC devices should be scrutinized, particularly hydrogen production and space propulsion.

1.2 Overview An excellent balance between potential reactor attractiveness and technical development risk motivates the study of field-reversed configuration (FRC) power plants. The linear, cylindrical FRC geometry facilitates the design of tritium-breeding blankets, shields, magnets, and input-power systems, while the high FRC beta increases the plasma power density and allows a compact fusion core. The surface heat flux is moderate despite a high power density, however, because the plasma flowing to the end chamber walls carries much of the fusion power.

Section 2: Fusion Core Engineering

2 Fusion Core Engineering

2.1 The thermo-mechanical design of the first wall and blanket of the FRC A helium-cooled solid breeder (Li2O) has been chosen for the FRC first wall and blanket. Oxide-dispersion strengthened (ODS) ferritic steel (developed at Oak Ridge National Lab (ORNL)) is the advanced structural material considered for the reactor components. This new material is exceptionally creep-resistant compared with low activation ferritic-martensitic (FM) steels at temperatures above 600°C.

2.2 The mechanical design of the first wall and blanket of the FRC The first zone and the outward consecutive layers of the blanket are made of concentric cylinders with the plasma in the center. The first zone consists of a first wall and a back wall made of steel cooled with helium and the space between them is filled with Li2O tubes also cooled with helium. All the tubes run longitudinally and have a circular cross section. The inner diameter of the first wall is 4.0 m. The outer diameter of the shield is 8.08 m.

2.3 Coolant routing To maximize the power conversion thermal efficiency the outlet helium temperature must be at the maximum attainable value. To achieve maximum power conversion thermal efficiency without violating all the constraints on the reactor materials’ maximum operating temperature, the helium coolant routing must be optimized. The route of the He gas coolant is as follows:

Cold He (T = 380°C) first enters all steel walls (first wall, blanket walls, and shield (steel)) to keep their temperature below 650°C.
Then He gas enters (T = 530°C) the Li2O zones (first zone, blanket-I, and blanket-II) to remove the generated volumetric heating.
The hot helium exits the blanket to the heat exchanger at about 830°C. The secondary helium exits the heat exchanger at about 800°C.

2.4 Thermal hydraulics calculations The total heating/module (volumetric and surface) in the first wall = 79.6 MW. Gas pressure is 18 MPa. Helium gas flow temperature rise in the first wall is 200°C.

2.7 Maintenance The special geometry (cylindrical) facilitates reasonable, practical, maintenance schemes that minimize the downtime and cost. The modular design allows the movement of the individual modules in the axial direction. To keep the vacuum integrity inside the reactor during operation a pillow type of overlap is used between modules.

Section 3: WISC: the WIsconsin Systems Code

3 WISC: the WIsconsin Systems Code

3.1 Introduction The WIsconsin Systems Code, WISC, is a C code for calculating field-reversed configuration (FRC) and tokamak power plant physics and engineering. An input file provides the data necessary for the code to calculate global power balance to determine the injection power necessary to sustain the plasma. The electron power balance is solved to obtain a self-consistent electron temperature. The power losses are due to charged particle transport, neutrons, and bremsstrahlung radiation. The code provides three energy conversion processes: direct conversion to electricity, thermal conversion, and conversion by both methods simultaneously.

The code accounts for six ion species: protons, deuterium (D), tritium (T), 3He, 4He, and an impurity. It determines the density of these species as a fraction of the deuterium density. The density of deuterium is based on either the beta limit or an input fixed averaged fuel density. The total ion pressure includes a contribution from a slowing-down distribution of fusion products.

Sections 3.3.1 through 3.3.27 specify the physics equations, Coulomb logarithms, FRC profiles, fusion reactivity expressions, fast ion slowing down, energy and particle confinement scalings, and power flow equations used in the model.

Section 4: Reference Case and APEX Liquid-Wall Cases

4 Reference Case and APEX Liquid-Wall Cases

4.1 Reference case Table 4-1 & 4-2 Reference Parameters Summary:

First wall radius: 2 m
Separatrix radius: 1.87 m
Separatrix length: 20 m
Core plasma volume: 220 m3
First-wall area: 251 m2
Average ion temperature: 24 keV
Average ion density: 1.5x10^20 m^-3
Volume-averaged beta: 0.56
Energy confinement time: 1.1 s
Vacuum magnetic field: 2.4 T
Blanket thickness: 1.45 m
Shield thickness: 0.61 m
Neutron wall load: 5.7 MW/m2
Neutron power: 1427 MW
Fusion power: 1785 MW
Total thermal power: 2114 MW
Thermal conversion efficiency: 0.52
Gross electric power: 1099 MWe
Net electric power: 1000 MWe
Mass power density: 430 kWe/Mg

4.2 APEX case comparison: Comparison between the solid wall reference case and the liquid-wall APEX case highlights significant increases in power density for the liquid wall concept (e.g., neutron wall load of 18 MW/m2 vs 5.7 MW/m2, mass power density of 1390 kWe/Mg vs 430 kWe/Mg).

5 Appendix: Key Papers Related to This Work

J.F. Santarius, G.A. Emmert, H.Y. Khater, E.A. Mogahed, C.N. Nguyen, L.C. Steinhauer, and G.H. Miley, “Field-Reversed Configuration Power Plant Critical Issues,” University of Wisconsin Fusion Technology Institute Report UWFDM-1084 (June 1998).
E.A. Mogahed, H.Y. Khater, and J.F. Santarius, “A Helium Cooled Li2O Pebble Bed Blanket Design for Cylindrical Geometry,” (prepared for ANS Topical Meeting on the Technology of Fusion Energy; thereby submitted to Fusion Technology, 2000).
R. W. Moir, T. D. Rognlien, K. Gulec, P. Fogarty, B. Nelson, M. Ohnishi, M. Rensink, J. F. Santarius, D. K. Sze, “Thick Liquid-walled, Field-reversed Configuration (FRC),” (prepared for ANS Topical Meeting on the Technology of Fusion Energy; thereby submitted to Fusion Technology, 2000).