M.Z. Hasan

The ARIES tokamak fusion reactor study

The Advanced Reactor Innovation and Evaluation Study (ARIES) is a community effort to develop several visions of the tokamak as a fusion power reactor. The aims are to determine its potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. The authors focus on the ARIES-1 design. Parametric systems studies show that the optimum first stability tokamak has relatively low plasma current ( approximately 12 MA), high plasma aspect ratio ( approximately 4-6), and high magnetic field ( approximately 24 T at the coil). ARIES-I is a 1000-MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m/sup 2/, and a mass power density of about 90 kWe/tonne. The ARIES-I reactor operates at steady state using ICRF (ion-cyclotron range of frequency) fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The ARIES-I blanket is cooled by He and consists of SiC-composite structural material, Li/sub 4/SiO/sub 4/ solid breeder, and Be neutron multiplier, all chosen for their low-activation and low-decay after-heat in order to enhance the safety and environmental features of the design. The ARIES-I design has a competitive cost of electricity and superior safety and environmental features.<<ETX>>

Blanket design for the ARIES-I tokamak reactor

For the Advanced Reactor Innovation and Evaluation Study-I (ARIES-I) tokamak power reactor design, the authors evaluated two gas-cooled, low-activation ceramic blanket designs, a 5-MPa helium-cooled design, and a 0.5-MPa CO/sub 2/ gas-carried, Li/sub 4/SiO/sub 4/ particulate design. The more extensive database available for the helium-cooled option has prompted the selection of this option as the reference design. The selected ARIES-I blanket design uses SiC composite as the structural material, 5-MPa helium as coolant, Li/sub 4/SiO/sub 4/ as the solid tritium breeder, and Be metal pellets as the neutron multiplier. This combination of materials provides the design of a high nuclear performance blanket with high-outlet temperature, good neutron multiplication, and adequate tritium breeding. It is a low-activation design that satisfies the criteria for 10CFR61 Class-C shallow land waste disposal, and achieves inherent safety since it produces negligible after heat, thus virtually eliminating the possibility of exposing the public to radioactivity. The mechanical design, neutronics analysis, thermal-hydraulic analysis, power-conversion system design, tritium extraction, and safety evaluation are summarized.<<ETX>>

Effects of Nonuniform Surface Heat Flux and Uniform Volumetric Heating on Blanket Design for Fusion Reactors

An analytical solution for the temperature profile and film temperature drop for fully-developed, laminar flow in a circular tube is provided. The surface heat flux varies circcimferentally but is constant along the axis of the tube. The volulmetric heat generation is uniform in the fluid. The fully developed laminar velocity profile is approximated by a power velocity profile to represent the flattening effect of a perpendicular magnetic field when the coolant is electrivally conductive. The presence of volumetric heat generation in the fluid adds another component to the film temperature drop to that due to the surface heat flux. The reduction of the boundary layer thickness by a perpendicular magnetic field reduces both of these two film temperature drops. A strong perpendicular magnetic field can reduce the film termperatiure drop by a factor of two if the fluid is electrically conducting. The effect of perpendicualr magnetic field )or the flatness of the velocity profile) is less pronounced on teh film termperature drop due to nonuniform surfacae heat flux than on that due to uniform surface heat flux. An example is provided to show the relative effects on these two film temperd

Introduction and synopsis of the TITAN reversed-field-pinch fusion-reactor study

Abstract The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density: and to determine the potential economic (cost of electricity), operational (maintenance and availability), safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs TITAN-I and TITAN-II, have been produced to demonstrate the possibility of multiple engineering-design approaches to high-MPD reactors. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the U.S. criteria for the near-surface disposal of radioactive waste (Class C, IOCFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a “single-piece” FPC maintenance procedure has been worked out and appears feasible for both designs. Parametric system studies have been used to find cost-optimized designs. to determine the parametric design window associated with each approach, and to assess the sensitivity of the designs to a wide range of physics and engineering requirements and assumptions. The design window for such compact RFP reactors would include machines with neutron wall loadings in the range of 10–20 MW/m 2 with a shallow minimum COE at about 18 MW/m 2 . Even though operation at the lower end of the this range of wall loading (10–12 MW/m 2 ) is possible, and may be preferable, the TITAN study adopted the design point at the upper end (18 MW/m 2 ) in order to quantify and assess the technical feasibility and physics limits for such high-MPD reactors. From this work, key physics and engineering issues central to achieving reactors with the features of TITAN-I and TITAN-II have emerged.

Overview of the TITAN-I fusion-power core

The TITAN reactor is a compact (major radius of 3.9 m and plasma minor radius of 0.6 m), high neutron wall loading (~18 MW/m 2 ) fusion energy system based on the reversed-field pinch (RFP) confinement concept. The reactor thermal power is 2918 MWt resulting in net electric output of 960 MWe and a mass power density of 700 kWe/tonne. The TITAN-I fusion power core (FPC) is a lithium, self-cooled design with vanadium alloy (V-3Ti-1Si) structural material. The surface heat flux incident on the first wall is ~4.5 MW/m 2 . The magnetic field topology of the RFP is favorable for liquid metal cooling. In the TITAN-I design, the first wall and blanket consist of single pass, poloidal flow loops aligned with the dominant poloidal magnetic field. A unique feature of the TITAN-I design is the use of the integrated-blanket-coil (IBC) concept. With the IBC concept the poloidal flow lithium circuit is also the electrical conductor of the toroidal-field and divertor coils. Three dimensional neutronics analysis yields a tritium breeding ratio of 1.18 and a molten salt extraction technique is employed for the tritium extraction system. Almost every FPC component would qualify for Class C waste disposal. The compactness of the design allows the use of single-piece maintenance of the FPC. This maintenance procedure is expected to increase the plant availability. The entire FPC operates inside a vacuum tank, which is surrounded by an atmosphere of inert argon gas to impede the flow of air in the system in case of an accident. The top-side coolant supply and return virtually eliminate the possibility of a complete LOCA occurring in the FPC. The peak temperature during a LOFA is 991 °C.

Properties of concentrated aqueous lithium nitrate solutions and applications to fusion reactor design

Aqueous solutions of Li-containing compounds have been proposed to serve as the combined tritium breeding material and coolant for fusion reactors. The salt used for the TITAN-II reversed-field-pinch reactor design is LiNO3, which was chosen for its good neutronics properties, relatively good corrosion characteristics, and for its high solubility in water. An extensive literature survey has shown that the physical and thermal properties of high-temperature, concentrated aqueous LiNO3 solutions are markedly different from the pure water properties at similar conditions. These changes alter the heat transfer performance of the coolant, and the critical heat flux is estimated to rise for sub-cooled flow boiling, while the heat transfer coefficient for forced convection falls. Another important result is the elevation in boiling point, which may allow the operating pressure of the primary coolant to be reduced to a value below that of the secondary steam circuit, preventing leakage of the tritiated coolant into the steam circuit. Further research is needed into corrosion and radiolysis issues, but the available data imply that careful control of the coolant chemistry can minimize the problems.

Thermal cycle power conversion for the ARIES-I tokamak reactor

The selection and analysis of a suitable power-conversion system for the Advanced Reactor Innovation and Evaluation Study-I (ARIES-I) tokamak fusion reactor are presented. Two main groups of thermal cycles have been investigated. One group is comprised of the nonconventional Brayton and Rankine dissociating-gas cycles using such reacting gases as nitrogen tetroxide (N/sub 2/O/sub 4/) and nitrosyl chloride (NOCl). The other group consists of the conventional inert-gas Brayton cycle and Rankine steam cycles. The dissociating-gas cycles have the potential to operate at higher temperatures and offer higher conversion efficiency and more compact design. However, because of the severe safety problems associated with the toxicity of N/sub 2/O/sub 4/ and NOCl, a dissociating-gas cycle was not selected. An inert-gas Brayton cycle, although more compact, offers much lower conversion efficiency compared to the Rankine steam cycle. The selected power-conversion system for ARIES-I is, therefore, based on advanced supercritical Rankine steam cycle with double reheat. A gross conversion efficiency of about 48% is predicted.<<ETX>>

Using the shield for thermal energy storage in pulsar

Abstract The PULSAR pulsed tokamak power plant design utilizes the outboard shield for thermal energy storage to maintain full 1000 MW(e) output during the dwell period of 200 s. Thermal energy resulting from direct nuclear heating is accumulated in the shield during the 7200 s fusion power production phase. The maximum shield temperature may be much higher than that for the blanket because radiation damage is significantly reduced. During the dwell period, thermal power discharged from the shield and coolant temperature are simultaneously regulated by controlling the coolant mass flow rate at the shield inlet. This is facilitated by throttled coolant bypass. Design concepts using helium and lithium coolant have been developed. Two-dimensional time-dependent thermal hydraulic calculations were performed to confirm performance capabilities required of the design concepts. The results indicate that the system design and performance can accommodate uncertainties in material limits or the length of the dwell period.

Thermal-hydraulic and structural design of the TITAN-I reversed-field-pinch fusion power core

Abstract Thermal-hydraulic and structural design of the first wall, blanket, and shield of the deuterium-tritium fueled TITAN-I reversed-field-pinch (RFP) fusion reactor is presented. Taking advantage of the characteristic low toroidal magnetic field of an RFP reactor, liquid lithium is used as the primary coolant to remove the thermal energy at an elevated temperature, thereby realizing a high power conversion efficiency of 44%. The use of liquid lithium has also led to a self-cooled design of the fusion power core in which the primary coolant is also the tritium breeder. The structural material is the vanadium alloy, V3Ti1Si. Tubular coolant channels are used in the first wall/blanket and rectangular channels in the hot shield. These are laid along the much larger poloidal field to minimize magnetohydrodynamic (MHD) pressure drop. Although the neutron wall loading of 18.1 MW/m2 is high, resulting in a radiation heat flux on the first wall of 4.6 MW/m2, three aspects of the design have made the removal of the reactor power at high temperature possible. These are: (1) the use of small-diameter circular tubes as coolant channels in the first wall, (2) the use of high-velocity MHD turbulent flow in the first-wall coolant tubes, and (3) thermal separation of the first-wall and blanket/shield coolant circuits, thereby allowing different exit temperatures. The thermal-hydraulic design was optimized by a design code developed for this purpose. Detailed structural design was performed by the finite element code ANSYS. The coolant inlet temperature is 320°C, and the coolant exit temperatures for the first-wall and blanket/shield coolant circuits are 442°C and 700°C, respectively. Lithium flow velocity in the first-wall coolant tubes is 21.6 m/s, and is ≤ 50 cm/s in the blanket/shield coolant channels. The total pressure drop in the first-wall coolant circuit is 10 MPa and in the blanket coolant circuit it is 3 MPa. The pumping power for coolant circulation is less than 5% of the net electric output. The material stresses are well within the design limits. The TITAN-I design suggests the feasibility and advantage of liquid-metal cooling of high wall loading RFP fusion reactors.

Heat Transfer in Plasma-Facing Components of Fusion Reactors: Non-MHD Laminar Flow in Rectangular Channels

AbstractConvective heat transfer in non-MHD laminar flow through rectangular channels in the first wall and limiter/divertor plates of fusion reactors has been analyzed numerically. Even for uniform heat flux, the Nusselt number (Nu) is not constant along the face of a rectangular channel, because the velocity is much smaller near a corner. For uniform heat flux, Nu varies by 67% from the center of a side to the corner (6.7 to 2.2). Therefore, the corners of a rectangular channel are possible hot-spot areas of concern for thermal-hydraulic designs. In addition, the surface heat flux on coolant channels in the plasma-facing components varies circumferentially. This nonuniformity of surface heat flux also affects the Nu. At the center of a side, Nu can be reduced from 6.7 to 2.8, i.e. by about 58%. For large nonuniformity of surface heat flux, the Nu at some locations can become infinity or negative; infinity, when the coolant/wall interface temperature becomes equal to the coolant bulk temperature and, neg...