S.J. Zinkle

Structural materials for fission & fusion energy

Structural materials represent the key for containment of nuclear fuel and fission products as well as reliable and thermodynamically efficient production of electrical energy from nuclear reactors. Similarly, high-performance structural materials will be critical for the future success of proposed fusion energy reactors, which will subject the structures to unprecedented fluxes of high-energy neutrons along with intense thermomechanical stresses. Advanced materials can enable improved reactor performance via increased safety margins and design flexibility, in particular by providing increased strength, thermal creep resistance and superior corrosion and neutron radiation damage resistance. In many cases, a key strategy for designing high-performance radiation-resistant materials is based on the introduction of a high, uniform density of nanoscale particles that simultaneously provide good high temperature strength and neutron radiation damage resistance.

Defect production in ceramics

A review is given of several important defect production and accumulation parameters for irradiated ceramics. Materials covered in this review include alumina, magnesia, spinel, silicon carbide, silicon nitride, aluminum nitride and diamond. Whereas threshold displacement energies for many ceramics are known within a reasonable level of uncertainty (with notable exceptions being AIN and Si3N4), relatively little information exists on the equally important parameters of surviving defect fraction (defect production efficiency) and point defect migration energies for most ceramics. Very little fundamental displacement damage information is available for nitride ceramics. The role of subthreshold irradiation on defect migration and microstructural evolution is also briefly discussed.

Operating temperature windows for fusion reactor structural materials

Abstract A critical analysis is presented of the operating temperature windows for nine candidate fusion reactor structural materials: four reduced-activation structural materials (oxide-dispersion-strengthened and ferritic/martensitic steels containing 8–12%Cr, V–4Cr–4Ti, and SiC/SiC composites), copper-base alloys (CuNiBe), tantalum-base alloys (e.g. Ta–8W–2Hf), niobium alloys (Nb–1Zr), and molybdenum and tungsten alloys. The results are compared with the operating temperature limits for Type 316 austenitic stainless steel. Several factors define the allowable operating temperature window for structural alloys in a fusion reactor. The lower operating temperature limit in all body-centered cubic (BCC) and most face-centered cubic (FCC) alloys is determined by radiation embrittlement (decrease in fracture toughness), which is generally most pronounced for irradiation temperatures below ∼0.3 T M where T M is the melting temperature. The lower operating temperature limit for SiC/SiC composites will likely be determined by radiation-induced thermal conductivity degradation, which becomes more pronounced in ceramics with decreasing temperature. The upper operating temperature limit of structural materials is determined by one of four factors, all of which become more pronounced with increasing exposure time: (1) thermal creep (grain boundary sliding or matrix diffusional creep); (2) high temperature He embrittlement of grain boundaries; (3) cavity swelling (particularly important for SiC and Cu alloys); or (4) coolant compatibility/corrosion issues. In many cases, the upper temperature limit will be determined by coolant corrosion/compatibility rather than by thermal creep or radiation effects. The compatibility of the structural materials with Li, Pb–Li, Sn–Li, He and Flibe (Li 2 BeF 4 ) coolants is summarized.

Radiation effects in ceramics

Abstract Ceramics represent a large class of solids with a wide spectrum of applicability, whose structures range from simple to complex, whose bonding runs from highly ionic to almost entirely covalent and, in some cases, partially metallic, and whose band structures yield wide-gap insulators, narrow-gap semiconductors or even superconductors. These solids exhibit responses to irradiation which are more complex than those for metals. In ceramic materials, atomic displacements can be produced by direct momentum transfer to often more than one distinguishable sublattice, and in some cases radiolytically by electronic excitations, and result in point defects which are in general not simple. Radiation-induced defect interaction, accumulation and aggregation modes differ significantly from those found in metals. Amorphization is a frequent option in response to high-density defect perturbation and is strongly related to structural topology. These fundamental responses to irradiation result in significant changes to important applicable properties, such as strength, toughness, electrical and thermal conductivities, dielectric response and optical behavior. The understanding of such phenomena is less well-understood than the simple responses of metals but is being increasingly driven by critical applications in fusion energy production, nuclear waste disposal and optical communications.

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.

Defect accumulation in pure fcc metals in the transient regime: A review

Abstract Over the years, a considerable amount of experimental results have been reported on defect accumulation in low-dose neutron irradiated pure fcc metals. In an effort to further the understanding of the processes involved in the defect accumulation in the transient regime, the experimental results are compiled and the salient features of these results are pointed out. Experimental results in pure aluminium, copper and nickel are chosen for this review. The dose dependence of the experimentally measured parameters makes it abundantly clear that the rate of build-up of cluster density, cavity density and the void swelling reaches a maximum at very low doses (≤ 0.1 dpa). There is no experimental evidence for the formation of a well defined dislocation network during irradiation of pure metals at temperatures in the void swelling regime up to a dose level of ~ 1 dpa. The experimental results allow us to identify three significant aspects of the defect accumulation behaviour under cascade damage conditions: (a) evolution of cavity microstructure in a spatially heterogeneous and segregated fashion, (b) high swelling rates at very low doses when the dislocation density is negligibly low and (c) enhanced vacancy accumulation in the vicinity of grain or subgrain boundaries. It is pointed out that these features cannot be rationalized in terms of conventional mean-field approach using chemical rate equations and dislocation bias as the only driving force. These features can be explained, however, by taking into considerations the specific nature of the cascade damage, namely, the intracascade recombination and clustering of interstitials and vacancies during the cooling down phase of a multidisplacement cascade.

Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

Abstract Microstructural data on the evolution of the dislocation loop, cavity, and precipitate populations in neutron-irradiated austenitic stainless steels are reviewed in order to estimate the displacement damage levels needed to achieve the “steady state” condition. The microstructural data can be conveniently divided into two temperature regimes. In the low temperature regime (below about 300°C) the microstructure of austenitic stainless steels is dominated by “black spot” defect clusters and faulted interstitial dislocation loops. The dose needed to approach saturation of the loop and defect cluster densities is generally on the order of 1 displacement per atom (dpa) in this regime. In the high temperature regime (~ 300 to 700°C), cavities, precipitates, loops and network dislocations are all produced during irradiation; doses in excess of 10 dpa are generally required to approach a “steady state” microstructural condition. Due to complex interactions between the various microstructural components that form during irradiation, a secondary transient regime is typically observed in commercial stainless steels during irradiation at elevated temperatures. This slowly evolving secondary transient may extend to damage levels in excess of 50 dpa in typical 300-series stainless steels, and to > 100 dpa in radiation-resistant developmental steels. The detailed evolution of any given microstructural component in the high-temperature regime is sensitive to slight variations in numerous experimental variables, including heat-to-heat composition changes and neutron spectrum.

Vanadium alloys – overview and recent results

Abstract This paper reviews recent progress in research on vanadium alloys with emphasis on V–4Cr–4Ti as a reference composition. New high purity V–4Cr–4Ti ingots and products (NIFS-HEATs) were made. The improved purity of the alloys made a practical demonstration of enhanced feasibility of recycling as a method of handling after use in fusion reactors. Significant progress has been made in the understanding of physical metallurgy of V–4Cr–4Ti and effects of O, N and C on the alloy properties such as low and high temperature mechanical properties, welding properties and low temperature irradiation effects, by means of including the comparison of various large heats and model alloys with different impurity levels. The effects of other trace impurities on some of the properties are also discussed. Other current efforts to characterize V–4Cr–4Ti, to improve its properties and to explore advanced vanadium alloys are reviewed. Issues remaining for the future investigations are discussed.

Fusion materials science: Overview of challenges and recent progress

A brief review is given of fundamental materials science concepts important for development of structural materials for fusion energy systems. Particular attention is placed on displacement damage effects associated with the unique deuterium-tritium fusion environment. Recent examples of multiscale materials modeling results (closely coupled with experimental studies) are summarized. Fundamental differences in the behavior of body centered cubic versus face centered cubic crystal structures are highlighted. Finally, a brief overview is given of the high-performance reduced-activation materials being developed by fusion.

Overview of Materials Research for Fusion Reactors

Materials research for fusion reactors is overviewed from Japanese, EU and US perspectives. Emphasis is placed on programs and strategies for developing blanket structural materials, and recent highlights in research and development for reduced activation ferritic martensitic steels, vanadium alloys and SiC/SiC composites, and in mechanistic experimental and modeling studies. The common critical issue for the candidate materials is the effect of irradiation with helium production. For the qualification of materials up to the full lifetime of a DEMO and Power Plant reactors, an intense neutron source with relevant fusion neutron spectra is crucial. Elaborate use of the presently available irradiation devices will facilitate efficient and sound materials development within the required time scale.