E.E. Bloom

The challenge of developing structural materials for fusion power systems

Nuclear fusion can be one of the most attractive sources of energy from the viewpoint of safety and minimal environmental impact. Central in the goal of designing a safe, environmentally benign, and economically competitive fusion power system is the requirement for high performance, low activation materials. The general performance requirements for such materials have been defined and it is clear that materials developed for other applications (e.g. aerospace, nuclear fission, fossil energy systems) will not fully meet the needs of fusion. Advanced materials, with composition and microstructure tailored to yield properties that will satisfy the specific requirements of fusion must be developed. The international fusion programs have made significant progress towards this goal. Compositional requirements for low activation lead to a focus of development efforts on silicon carbide composites, vanadium alloys, and advanced martensitic steels as candidate structural material systems. Control of impurities will be critically important in actually achieving low activation but this appears possible. Neutron irradiation produces significant changes in the mechanical and physical properties of each of these material systems raising feasibility questions and design limitations. A focus of the research and development effort is to understand these effects, and through the development of specific compositions and microstructures, produce materials with improved and adequate performance. Other areas of research that are synergistic with the development of radiation resistant materials include fabrication, joining technology, chemical compatibility with coolants and tritium breeders and specific questions relating to the unique characteristics of a given material (e.g. coatings to reduce gas permeation in SiC composites) or design concept (e.g. electrical insulator coatings for liquid metal concepts).

International strategy for fusion materials development

Abstract In this paper, the results of an IEA-Workshop on Strategy and Planning of Fusion Materials Research and Development (R&D), held in October 1998 in Riso Denmark are summarised and further developed. Essential performance targets for materials to be used in first wall/breeding blanket components have been defined for the major materials groups under discussion: ferritic–martensitic steels, vanadium alloys and ceramic composites of the SiC/SiC-type. R&D strategies are proposed for their further development and qualification as reactor-relevant materials. The important role of existing irradiation facilities (mainly fission reactors) for materials testing within the next decade is described, and the limits for the transfer of results from such simulation experiments to fusion-relevant conditions are addressed. The importance of a fusion-relevant high-intensity neutron source for the development of structural as well as breeding and special purpose materials is elaborated and the reasons for the selection of an accelerator-driven D-Li-neutron source – the International Fusion Materials Irradiation Facility (IFMIF) – as an appropriate test bed are explained. Finally the necessity to execute the materials programme for fusion in close international collaboration, presently promoted by the International Energy Agency, IEA is emphasised.

Low activation materials for fusion applications

Abstract In February 1982, the Office of Fusion Energy of the U.S. Department of Energy established a panel to examine the possibility of developing and using materials with attractive radioactivation characteristics for applications in fusion power reactors. This paper summarizes the findings of the panel.

The effects of large concentrations of helium on the mechanical properties of neutron-irradiated stainless steel☆

Abstract Tensile and creep properties have been determined on specimens of type 316 stainless steel irradiated in the High Flux Isotope Reactor in the range 380 to 785°C. Irradiation of type 316 in this reactor partially simulates fusion reactor irradiation, with displacement damage levels up to 120 dpa and helium contents up to 6000 appm achieved in two years. Samples irradiated in the annealed condition to about 100 dpa and 4000 appm helium showed an increased yield strength between 350 and 600°C and, except at 350°C, a reduced ultimate tensile strength compared with values for the unirradiated material. Samples irradiated in the 20%-cold-worked condition showed decreases in both yield and ultimate tensile strengths at all test temperatures. The irradiated samples of both annealed and cold-worked material exhibited little strain hardening, and total elongations were small and became zero,for tests at 650° C. Tensile tests at 575°C and creep-rupture tests at 550°C showed strong effects of fluence on strength and ductility for helium contents above about 30 appm. Optical metallography showed extensive carbide precipitation at all temperatures and precipitation of a second phase, believed to be sigma, at the higher temperatures.

The effect of minor constituents on swelling in stainless steel

Abstract The small quantities of solute interstitial elements in stainless steel (C, N and possibly Si) reduce the swelling under neutron irradiation ( ∼ 2 × 10 22 neutrons/cm 2 ) by more than an order of magnitude between 500 and 600° C over high purity material. The solute interstitials reduce both the numbers and sizes of irradiation-caused voids. Current swelling models ignore — of necessity — this gross effect. Several possible mechanisms are suggested to account for the effect.

Mechanical properties of materials in fusion reactor first-wall and blanket systems

With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n, α) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above 0.5 Tm. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their higher melting temperatures.

Impurity effects on reduced-activation ferritic steels developed for fusion applications

Abstract Reduced-activation steels are being developed for fusion applications by restricting alloying elements that produce long-lived radioactive isotopes when irradiated in the fusion neutron environment. Another source of long-lived isotopes is the impurities in the steel. To examine this, three heats of reduced-activation martensitic steel were analyzed by inductively coupled plasma mass spectrometry for low-level impurities that compromise the reduced-activation characteristics: a 5-ton heat of modified F82H (F82H-Mod) for which an effort was made during production to reduce detrimental impurities, a 1-ton heat of JLF-1, and an 18-kg heat of ORNL 9Cr–2WVTa. Specimens from commercial heats of modified 9Cr–1Mo and Sandvik HT9 were also analyzed. The objective was to determine the difference in the impurity levels in the F82H-Mod and steels for which less effort was used to ensure purity. Silver, molybdenum, and niobium were found to be the tramp impurities of most importance. The F82H-Mod had the lowest levels, but in some cases the levels were not much different from the other heats. The impurity levels in the F82H-Mod produced with present technology did not achieve the low-activation limits for either shallow land burial or recycling. The results indicate the progress that has been made and what still must be done before the reduced-activation criteria can be achieved.

The effects of large fast-neutron fluences on the structure of stainless steel☆

Abstract The changes in microstructure which occur when type 304L stainless steel is irradiated at temperatures between 370 and 472 °C and fast-neutron fluences between 0.8 and1.4 × 1022n/cm2 have been investigated by electron microscopy. The fuel cladding from an Experimental Breeder Reactor-II driver fuel element served as the experimental material. Voids and dislocation loops formed by the precipitation of irradiation-produced vacancies and interstitials were observed in each specimen. As the irradiation temperature was increased the density of each type of defect decreased and the average size increased. Upon postirradiation annealing at 600 °C, the dislocation loops converted to a dislocation network and the smallest voids ( diameters A ) began to disappear. As the annealing temperature was increased the dislocation and void densities both decreased. After annealing 1 h at 900 °C, the voids were completely removed and the dislocation density was reduced to that of an unirradiated annealed specimen.

Austenitic stainless steels with improved resistance to radiation-induced swelling☆

Stainless steel specimens were bombarded with Ni ions at elevated temperatures to simulate fast neutron damage. Results show that type 316 stainless steel with additions of silicon and titanium exhibits low swelling over the entire swelling temperature range under high-dose nickel-ion bombardment. Neutron irradiation data on commercial alloys and ion data on high-purity alloys also indicate that the most effective suppression of swelling is achieved with combinations of silicon and titanium. It is suggested that suppression of swelling by alloying with silicon and titanium may be effective over a range of nickel and chromium base composition levels and will provide the basis for the development of low-swelling alloys that are technologically similar to type 316 stainless steel. It should be noted that the influence of silicon and titanium on swelling is likely to depend strongly on the concentrations of other elements such as carbon, oxygen, and nitrogen and on the extent to which silicon and titanium partition to various phases.

The effect of irradiation temperature on strength and microstructure of stainless steel

Abstract The effects of irradiation at temperatures between 93 and 454° C upon the room-temperature mechanical properties and electron microstructuros of AISI type 304 stainless steel have been determined. Irradiation at temperatures between 93 and 300° C produced a high density of defect clusters on the order of 100 A in dia. which wore responsible for an increased yield stress. After irradiation at 371°C practically no defect clusters were observed and the yield stress was lower by a factor of two. At irradiation temperatures of 371° C and higher, precipitates formed within the grains. Deformation (10% by rolling) in a specimen containing the defect clusters was concentrated in very narrow slip bands, while in a specimen containing precipitate particles, the deformation was homogeneous.