M.C. Billone

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.

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

Fusion power core engineering for the ARIES-ST power plant

Abstract ARIES-ST is a 1000 MWe fusion power plant based on a low aspect ratio ‘spherical torus’ (ST) plasma. The ARIES-ST power core was designed to accommodate the unique features of an ST power plant, to meet the top-level requirements of an attractive fusion energy source, and to minimize extrapolation from the fusion technology database under development throughout the world. The result is an advanced helium-cooled ferritic steel blanket with flowing PbLi breeder and tungsten plasma-interactive components. Design improvements, such as the use of SiC inserts in the blanket to extend the outlet coolant temperature range were explored and the results are reported here. In the final design point, the power and particle loads found in ARIES-ST are relatively similar to other advanced tokamak power plants (e.g. ARIES-RS [Fusion Eng. Des. 38 (1997) 3; Fusion Eng. Des. 38 (1997) 87]) such that exotic technologies were not required in order to satisfy all of the design criteria. Najmabadi and the ARIES Team [Fusion Eng. Des. (this issue)] provide an overview of ARIES-ST design. In this article, the details of the power core design are presented together with analysis of the thermal–hydraulic, thermomechanical and materials behavior of in-vessel components. Detailed engineering analysis of ARIES-ST TF and PF systems, nuclear analysis, and safety are given in the companion papers [4] , [5] , [6] , [7] .

Radial-hydride Embrittlement of High-burnup Zircaloy-4 Fuel Cladding

Prestorage drying operations of high-burnup fuel may make Zircaloy-4 (Zry-4) fuel cladding more susceptible to failure, especially during fuel handling, transport, and post-storage retrieval. In particular, hydride precipitates may reorient from the circumferential to the radial direction of the cladding during drying operations if a threshold level of hoop stress at or above a corresponding threshold temperature is exceeded. This study indicates that the threshold stress is approximately 75–80 MPa for both nonirradiated and high-burnup stress-relieved Zry-4 fuel cladding cooled from 400°C and, under ring compression at both room temperature and 150°C, that radial-hydride precipitation embrittles Zry-4. Specifically, the plastic tensile hoop strain needed to initiate unstable crack propagation along radial hydrides decreases dramatically from >8% to lt;1% as radial-hydride fraction increases. Lower hydrogen contents (lr;300wppm) appear to be more susceptible to radial-hydride embrittlement compared to higher contents (>600 wppm), like that found in high-burnup Zry-4.

Diffusion/desorption of tritium from irradiated beryllium

This report discusses stepped-thermal-anneal tritium-release measurements at 573 up to 1173 K which have been performed on irradiated Be test materials, fabricated and irradiated to meet conditions relevant to the International Thermonuclear Experimental Reactor (ITER). A combined diffusion/desorption model for tritium release allows determination of diffusion coefficients and desorption-rate constants in the mixed-mechanism regime where both diffusion and surface desorption appear to be rate-limiting. The effective tritium diffusivities (m{sup 2}/s) for these materials, and also from new data analysis of previously reported fully dense material, were found to be: 81% TD Be:1.7 {times} 10{sup {minus}11} exp({minus}3.5 kJ/mol/RT); 99% TD Be:1.6 {times} 10{sup {minus}10} exp({minus}9.5 kJ/mol/RT); 100% TD Be: 1.4 {times} 10{sup {minus}10} exp({minus}11.5 kJ/mol/RT). Tritium release data for both the 81% TD and the 99% TD sample were matched reasonably well by the diffusion/desorption model. The model provides evidence for a changing mechanism over both temperature and density, but indicates that diffusion is the primary mechanism with a small and changing contribution from surface desorption.

The trio experiment

The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an analytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion.

Modeling of tritium release from ceramic breeders: Status and some implications for next-step devices

Abstract As the International Thermonuclear Experimental Reactor (ITER) joint activities are approaching the phase of detailed engineering design, it is crucial to survey the status of development of theoretical models and computational methodologies for tritium transport in, and release from, fusion ceramic breeders anticipated to be irradiated in the next-step device. This paper reviews the tritium modeling area and summarizes recent progress made in the development of more accurate predictive capabilities. The areas upon which attention is focused include: identification of mechanisms and rate-limiting steps; material data-base status and needs; model validation and comparison with experimental results; parameter ranges of applicability; areas of latest advance and remaining crucial issues which require additional effort. Current model applications such as analysis of experimental results available from small-scale in-pile experiments, and tritium inventory calculations for fusion blankets of conceptual designs are discussed. Areas of future applications are identified and emphasis is placed upon the growing need to improve on existing predictive capabilities to interpret tritium behavior and release results from solid breeder test components under pulsed conditions in ITER as well as to estimate more rigorously blanket tritium inventories.

Tritium and helium behavior in irradiated beryllium

Large quantities of Be (> 100 metric tons) are planned for use in the ITER blanket design to enhance tritium breeding and to act as a thermal barrier between coolant and breeder. Tritium retention/release and He-induced swelling are important issues in blanket design. The data base on tritium and helium behavior in Be is reviewed. New data on tritium retention/release and He bubble growth are presented for Be irradiated to 5 {times} 10{sup 22} n(E > 1 MeV)/cm{sup 2} at {approximately}75{degree}C and postirradiation-annealed for 700 hours at 500{degree}C. A model (diffusion/desorption) is proposed and tested against the data base to determine tritium diffusivity and the desorption rate constant. Similarly a model for He-induced swelling is developed and tested against the data base. The dependence of tritium retention and release on He content and impurities (e.g. BeO) is also explored. 11 refs., 6 figs.

Materials integration issues for high performance fusion power systems

One of the primary requirements for the development of fusion as an energy source is the qualification of materials for the frost wall/blanket system that will provide high performance and exhibit favorable safety and environmental features. Both economic competitiveness and the environmental attractiveness of fusion will be strongly influenced by the materials constraints. A key aspect is the development of a compatible combination of materials for the various functions of structure, tritium breeding, coolant, neutron multiplication and other special requirements for a specific system. This paper presents an overview of key materials integration issues for high performance fusion power systems. Issues such as: chemical compatibility of structure and coolant, hydrogen/tritium interactions with the plasma facing/structure/breeder materials, thermomechanical constraints associated with coolant/structure, thermal-hydraulic requirements, and safety/environmental considerations from a systems viewpoint are presented. The major materials interactions for leading blanket concepts are discussed.

Status of beryllium development for fusion applications

Abstract Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma-facing component of first-wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold-isostatic-pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well as its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation. In this current work, the range of anticipated fusion operating conditions is reviewed. The thermal, mechanical, chemical compatibility, tritium retention/release, and helium retention/swelling databases are then reviewed for fabrication methods and fusion operating conditions of interest. Properties correlations and uncertainty ranges are also discussed. In the case of the more complex phenomena of tritium retention/release and helium-induced swelling, fundamental mechanisms and models are reviewed in more detail. Areas in which additional data are needed are highlighted, along with some trends which suggest ways of optimizing the performance of beryllium for fusion applications.