A.R. Raffray

Design and material issues for high performance SiCf/SiC-based fusion power cores

The SiCf/SiC composite is a promising structural material candidate for fusion power cores and has been considered internationally in several power plant studies. It offers safety advantages arising from its low induced radioactivity and afterheat, and the possibility of high performance through high temperature operation. However, its behavior and performance at high temperatures and under irradiation are still not well known and need to be better characterized. This paper summarizes the current SiCf/SiC design and R&D status. The latest SiCf/SiC-based power core design studies are summarized, and the key SiCf/SiC parameters affecting the performance of power core components are highlighted. The current status of the material R&D is discussed, with the focus on fabrication and joining, baseline properties and properties under irradiation, as well as the desirable evolution of these properties. In the light of this, the R&D plans are summarized and assessed. Finally, to help present-day design studies and in the expectation of future confirmatory R&D results, recommendations are provided on SiCf/SiC parameters and properties to be assumed for present design analysis of long term SiCf/SiC-based power plants.

Critical heat flux analysis and R&D for the design of the ITER divertor

The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW:m 2 over :10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R&D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R&D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of : 1.4 or more for all divertor components.

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.

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.

Mistral: A comprehensive model for tritium transport in lithium-base ceramics: Part I: Theory and description of model capabilities☆

Abstract MISTRAL is a theoretical model developed to describe tritium transport and release in fine-grained ceramic materials for tritium breeding applications in fusion blankets. The model includes as relevant physical processes tritium diffusion in the effective bulk (grains and grain boundaries), adsorption, recombination and desorption at the breeder surface and diffusion through the network of pores. A key improvement of the model, compared with those already in the literature, consists of a better characterization of the processes at the breeder surface and their linking to the bulk and pore regions. The sets of governing transport equations and corresponding boundary conditions are formulated together with the choice of the computational algorithm. A computer code with transient capabilities has been developed based on the model. It aims at describing tritium release for several transient conditions relevant for in-pile tritium recovery experiments and for fusion blankets. To assess the range of applicability of the model, several calculations have been performed and the results of the analysis are herein presented and discussed.

The ITER blanket system design challenge

This paper summarizes the latest progress in the ITER blanket system design as it proceeds through its final design phase with the Final Design Review planned for Spring 2013. The blanket design is constrained by demanding and sometime conflicting design and interface requirements from the plasma and systems such as the vacuum vessel, in-vessel coils and blanket manifolds. This represents a major design challenge, which is highlighted in this paper with examples of design solutions to accommodate some of the key interface and integration requirements.

The Science and Technologies for Fusion Energy With Lasers and Direct-Drive Targets

We are carrying out a multidisciplinary multi-institutional program to develop the scientific and technical basis for inertial fusion energy (IFE) based on laser drivers and direct-drive targets. The key components are developed as an integrated system, linking the science, technology, and final application of a 1000-MWe pure-fusion power plant. The science and technologies developed here are flexible enough to be applied to other size systems. The scientific justification for this work is a family of target designs (simulations) that show that direct drive has the potential to provide the high gains needed for a pure-fusion power plant. Two competing lasers are under development: the diode-pumped solid-state laser (DPPSL) and the electron-beam-pumped krypton fluoride (KrF) gas laser. This paper will present the current state of the art in the target designs and lasers, as well as the other IFE technologies required for energy, including final optics (grazing incidence and dielectrics), chambers, and target fabrication, injection, and tracking technologies. All of these are applicable to both laser systems and to other laser IFE-based concepts. However, in some of the higher performance target designs, the DPPSL will require more energy to reach the same yield as with the KrF laser.

Direct drive target survival during injection in an inertial fusion energy power plant

In inertial fusion energy (IFE) power plant designs, the fuel is a spherical layer of frozen DT contained in a target that is injected at high velocity into the reaction chamber. For direct drive, typically laser beams converge at the centre of the chamber (CC) to compress and heat the target to fusion conditions. To obtain the maximum energy yield from the fusion reaction, the frozen DT layer must be at about 18.5 K and the target must maintain a high degree of spherical symmetry and surface smoothness when it reaches the CC. During its transit in the chamber the cryogenic target is heated by radiation from the hot chamber wall. The target is also heated by convection as it passes through the rarefied fill-gas used to control chamber wall damage by x-rays and debris from the target explosion. This article addresses the temperature limits at the target surface beyond which target uniformity may be damaged. It concentrates on direct drive targets because fuel warm up during injection is not currently thought to be an issue for present indirect drive designs and chamber concepts. Detailed results of parametric radiative and convective heating calculations are presented for direct-drive targets during injection into a dry-wall reaction chamber. The baseline approach to target survival utilizes highly reflective targets along with a substantially lower chamber wall temperature and fill-gas pressure than previously assumed. Recently developed high-Z material coatings with high heat reflectivity are discussed and characterized. The article also presents alternate target protection methods that could be developed if targets with inherent survival features cannot be obtained within a reasonable time span.