S. Malang

On the exploration of innovative concepts for fusion chamber technology

Abstract This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load >10 MW/m 2 and surface heat flux >2 MW/m 2 , (2) high power conversion efficiency (>40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further research and development. The first class seeks to eliminate the solid “bare” first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn–Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin (∼2 cm) or thick (∼40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma β and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma–liquid interactions including both plasma–liquid surface and liquid wall–bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W–5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at ∼1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.

Overview of the ARIES-RS reversed-shear tokamak power plant study

The ARIES-RS tokamak is a conceptual, D‐T-burning 1000 MWe power plant. As with earlier ARIES design studies, the final design of ARIES-RS was obtained in a self-consistent manner using the best available physics and engineering models. Detailed analyses of individual systems together with system interfaces and interactions were incorporated into the ARIES systems code in order to assure self-consistency and to optimize towards the lowest cost system. The ARIES-RS design operates with a reversed-shear plasma and employs a moderate aspect ratio (A4.0). The plasma current is relatively low (Ip11.32 MA) and bootstrap current fraction is high ( fBC 0.88). Consequently, the auxiliary power required for RF current drive is relatively low ( 80 MW). At the same time, the average

Breeding blanket concepts for fusion and materials requirements

This paper summarizes the design and performances of recent breeding blanket concepts and identifies the key material issues associated with them. An assessment of different classes of concepts is carried out by balancing out the potential performance of the concepts with the risk associated with the required material development. Finally, an example strategy for blanket development is discussed.

Conceptual design of the dual-coolant blanket in the frame of the EU power plant conceptual study

Abstract The dual-coolant (DC) blanket—characterised by its simple construction, simple function, and high thermal efficiency—is one of the EU advanced blanket concepts to be investigated in the frame of the long-term power plant conceptual study (PPCS). Its basic concept is based on the use of helium-cooled ferritic steel structure, the self-cooled Pb–17Li breeding zone, and SiC/SiC flow channel inserts, serving as electrical and thermal insulators. The present work on PPCS is drawn extensively on the preparatory study on plant availability carried out in 1999 with an objective to perform the conceptual design of the DC blanket concept where some details are to be selected in accordance with the overall strategy, which allows an extrapolation of the present knowledge between the near-term solutions (helium-cooled pebble bed (HCPB), water-cooled lead–lithium (WCLL) blanket concepts), and the very advanced self-cooled Pb–17Li SiC/SiC (SCLL) blanket concept. In the PPCS the reactor power is adapted to a typical size of commercial reactors of 1500 MWe which requires iterative calculations between the blanket layout and the system code analysis. The results of the first iteration are reported. This work is under the coordination of FZK in co-operation with CEA, EFET, IBERTEF, UKAEA, VTT Processes and VR.

High performance PbLi blanket

A novel blanket concept is described. The proposed design Is a dual coolant concept based on ferritic steel as structural material using helium to cool the first wall. The temperature of the entire steel structure is maintained below the 550/spl deg/C limit. The breeding zone is cooled by circulating the liquid metal breeder to external heat exchangers. Flow channel inserts are employed in the poloidal liquid breeder ducts, serving both as electrical and thermal insulator between the flowing liquid metal and the steel structure. In this way, a liquid metal exit temperature of about 700/spl deg/C is achievable, allowing either an advanced Rankine steam cycle or closed-cycle helium gas turbine (Brayton cycle) as the power conversion system. A gross thermal efficiency of about 45% can be achieved with either system.

U.S. PLANS AND STRATEGY FOR ITER BLANKET TESTING

Abstract Testing blanket concepts in the integrated fusion environment is one of the principal objectives of ITER. Blanket test modules will be inserted in ITER from Day 1 of its operation and will provide the first experimental data on the feasibility of the D-T cycle for fusion. With the US rejoining ITER, the US community has decided to have strong participation in the ITER Test Blanket Module (TBM) Program. A US strategy for ITER-TBM has evolved that emphasizes international collaboration. A study was initiated to select the two blanket options for the US ITER-TBM in light of new R&D results from the US and world programs over the past decade. The study is led by the Plasma Chamber community in partnership with the Materials, PFC, Safety, and physics communities. The study focuses on assessment of the critical feasibility issues for candidate blanket concepts and it is strongly coupled to R&D of modeling and experiments. Examples of issues are MHD insulators, SiC insert viability and compatibility with PbLi, tritium permeation, MHD effects on heat transfer, solid breeder “temperature window” and thermomechanics, and chemistry control of molten salts. A dual coolant liquid breeder and a helium-cooled solid breeder blanket concept have been selected for the US ITER-TBM.

High performance blanket for ARIES-AT power plant

The ARIES-AT blanket has been developed with the overall objective of achieving high performance while maintaining attractive safety features, simple design geometry, credible maintenance and fabrication processes, and reasonable design margins as an indication of reliability. The design is based on Pb–17Li as breeder and coolant and SiCf/SiC composite as structural material. This paper summarizes the results of the design study of this blanket.

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.

The EU advanced dual coolant blanket concept

Abstract The advanced dual coolant (A-DC) blanket is one of the EU advanced concepts to be investigated in the frame of the long-term power plant conceptual study (PPCS). Its basic concept—following the ARIES-ST concept—is based on the use of helium-cooled ferritic steel structure, the self-cooled Pb–17Li breeding zone, and SiC/SiC flow channel inserts. The latter serves as electrical and thermal insulators and therefore minimize the pressure losses and enable a relatively high Pb–17Li exit temperature leading to a high thermal efficiency. The present work on PPCS is drawn extensively on the preparatory study on plant availability (PPA) carried out in 1999 where a maximum neutron wall load of 5 MW/m2 (corresponding maximum surface heat load of 0.9 MW/m2) was given in the reference case of the A-DC blanket. In the following stage of PPCS the A-DC blanket is normalized and adapted to a typical size of commercial reactors (e.g. 1500 MWe) which requires iterative calculations between the blanket layout and the system code analysis. The status of the work with some idea improvements is reported.