F.W. Wiffen

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 ductility in bending of molybdenum alloys irradiated between 425 and 1000°C☆

Abstract Molybdenum and the alloys Mo-0.5% Ti and TZM were irradiated at four temperatures between 425 and 1000°C to a displacement level of 11 dpa at a fluence of 2.5 × 10 26 n / m 2 (>0.1 MeV ). Vacuum bend tests at elevated temperatures showed the ductile-to-brittle transition temperature (DBTT) to be above room temperature for all irradiation temperatures. Irradiation at 585°C produced the highest DBTT, 550 to 700°C. Differences among the three materials were minor.

Special purpose materials for fusion application

Originally in 1978 the Special Purpose Materials Task Group was concerned with tritium breeding materials, coolants, tritium barriers, graphite and silicon carbide, ceramics, heat-sink materials, and magnet components. Since then several other task groups have been created, so that now the category includes only materials for superconducting magnets and ceramics. For the former application copper-stabilized Nb3Sn (Ti) insulated with polyimides will meet the general requirements, so that testing of prototype components is the priority task. Ceramics are required for several critical components of fusion reactors either as dielectrics or as a structural material. Components near the first wall will receive exposures of 5–20 MW y/m2. Other ceramic applications are well behind the first wall, with lower damage levels. Most insulators operate near room temperature, but ceramic blanket structures may operate up to 1000°C. Because of a meager data base, one cannot identify optimum ceramics for structural application; but MgAl2O4 is an attractive dielectric material.

Microstructures developed in ‘simulated’ fusion irradiations☆

Abstract Microstructural information is reviewed for irradiation conditions which ‘simulate’ one or more salient features of the irradiation environment anticipated at first wall and blanket positions in a fusion reactor. The effects of high energy neutron damage from T[d,n], Be[d,n] and 16 MeV proton irradiations, and the effects of simultaneous atom displacement damage and helium and hydrogen production from mixed-spectrum fission reactor and dual-ion irradiation experiments are summarized for a variety of metals. It is shown that observations of microstructural sensitivity to damage rate, irradiation temperature and helium are providing important clues for developing damage resistant structural alloys for fusion reactor appiications.

The microstructure and swelling of neutron irradiated tantalum

Electron microscopy of tantalum that was irradiated to 2.5 × 1022 neutrons/cm2(⪢0.1 MeV) at temperatures between 425 and 1050°C established that this temperature range effectively brackets the swelling temperatures of tantalum. No voids were detected after irradiation at 425°C. The temperature for maximum swelling was inferred to be near 600°C, and the swelling at ~1000°C was very small. At 585 and 790°C a high concentration of small voids resulted in technologically significant levels of swelling, 0.5 to 2.5%.

Tracing the evolution of bubbles in helium-injected aluminum by means of positron annihilation

Abstract Pure aluminum was injected with 0.6 appm of helium. Positron lifetime measurements were performed following isochronal annealing between 25 and 600° C. Transmission electron micrographs were taken after several of these anneals. The results are consistent with a model in which the injected helium quickly becomes substitutional. This substitutional helium is a trapping site for positrons. By monitoring the concentration and size of this trapping site through the measurement of the positron lifetimes it is concluded that migration of the helium begins at ~ 100°C and that helium continually agglomerates up to ~250°C. Above this temperature the increasing size of the trapping site indicates bubble growth. Bubble growth proceeds up to 600° C, but whether by continued arrival of helium at nucleation sites or by bubble migration and coalescence (or both) being indeterminable. The major stage of bubble growth is between 300 and 600° C. The electron microscopy verified the presence of bubbles both in grain boundaries and matrix after annealing to 625°C, but only cavities in grain boundaries were observed in annealing to 500° C. Positron-lifetime measurements are a useful complement to electron microscopy in the study of bubbles in metals, especially in the early stages of the nucleation and growth of small cavities.

Structural Materials Data Base Assessment for the Blanket Comparison and Selection Study

The Blanket Comparison and Selection Study (BCSS) had as its primary goal the selection of a limited number of blanket concepts for fusion power reactors, to serve as the focus for the U.S. Departm...

Materials engineering for TFCX

Four point designs for TFCX resulted from a US national effort that was completed in July 1984. The goal of these devices is to achieve D-T ignition and long pulse operation. They differ primarily in the assumptions for the toroidal field coils. Two use copper coils, the other two Nb3Sn superconducting coils. Differences in each pair derive from the assumed coil performance parameters, including stress levels, current density, and coil cooling. The bigger single materials question is the radiation limits on the coil insulation, but a large number of materials questions were identified by the study. The TFCX is reactor relevant in terms of neutron, particle and heat flux values: its limited design lifetime makes fluence levels far short of levels projected for power reactors. These and other materials issues raised by this project are discussed.

Copper and copper alloys for fusion reactor applications: Summary Report of a DOE-OFE workshop☆

Abstract The Office of Fusion Energy of the U.S. Department of Energy convened a workshop in April 1983 to review the needs for copper and copper alloys in fusion device applications. The adequacy of the data base on these materials was examined, and recommendations were developed for experimental programs needed to fill identified data gaps. The workshop results are available in a conference proceedings.