It is the general conclusion of all national programs that the development of high-performance reduced-activation structural materials is essential for the successful development of fusion power. In this paper, the experience gleaned from previous programs to develop materials for high temperature structural applications is used to identify and discuss some of the most critical issues that must be addressed in the development of candidate materials for fusion structural applications. Critical issues discussed include radiation-induced solute segregation and implications on phase stability in the development of high-performance alloys/ceramics; the effects of very large amounts of helium on mechanical properties and the implications for alloy design/development; development of high temperature design methodology and incorporation of radiation effects into this methodology; the effects of radiation damage on flow localization, and the implications and approach to control the phenomena; and considerations of mass transfer and corrosion in complex fusion systems.

High-performance reduced-activation materials are crucial for fulfillment of the promise of fusion to provide safe, economical, and environmentally acceptable energy. Three reduced activation structural materials have emerged as promising candidates, based on 8–9Cr ferritic/martensitic steels, V–Cr–Ti alloys, and SiC/SiC composites. Due to advances in understanding how to control and engineer the nanoscale phase stability required for harsh neutron irradiation environments, these reduced activation materials have unirradiated properties that are superior to commercially available analogs. Perhaps the most important accomplishment to date from fusion materials research is the radiation effects knowledge base. Models of radiation effects and supporting experiments highlight the critical role of helium production on the microstructural stability and lifetime of irradiated materials. The proposed International Fusion Materials Irradiation Facility (IFMIF) would fill a critical need for fusion materials development.

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 Risø 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.

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.

Neutron irradiation at low temperatures (100 to 275 degrees C) to 0.5 displacements per atom causes significant embrittlement and changes in the subsequent room temperature tensile properties of V-4Cr-4Ti. The yield strength and microhardness at room temperature increase with increasing irradiation temperature. The tensile properties at room temperature show large increases in strength and a complete loss of work hardening capacity with no uniform elongation. Embrittlement, as measured by an increase in the ductile-to-brittle transition temperature, increases with increasing irradiation temperature, at least up to 275 degrees C. It is shown that this embrittlement is not due to pickup of O or other interstitial solutes during the irradiation.

In the present work, the formation of amorphous zones in 6H-SiC and 3C-SiC by low fluence (4×108 to 1×1012 cm−2) 420 keV Xe implantation at room temperature is investigated with positron beam analysis (PBA). By relating the fraction of positrons annihilating in amorphous zones obtained by PBA to stopping and range of ions in matter calculations in combination with a local damage level criterion for amorphization, the amorphous zones are characterized. Furthermore, the recovery by annealing of 6H-SiC after implantation with 2 different ion fluences (2×1010 and 2×1011 cm−2) is studied with PBA and a model assuming shrinkage of amorphous zones.

It has long been recognized that attainment of the safety and environmental potential of fusion energy requires the successful development of low activation materials for the first wall, blanket and other high heat flux structural components. Only a limited number of materials potentially possess the physical, mechanical and low activation characteristics required for this application. The current US structural materials research effort is focused on three candidate materials: advanced ferritic steels, vanadium alloys and silicon carbide composites. Recent progress has been made in understanding the response of these materials to neutron irradiation.

This paper will present a brief overview of the present understanding of radiation effects in materials. Progress in fundamental understanding of void swelling, phase formation and stability, irradiation creep, and post-irradiation mechanical properties has been gained through comparison of experimental data with the predictions of theoretical models. Applications of the understanding of radiation effects in materials have led to the development of candidate structural materials for use in aggressive radiation environments. Compositional and microstructural controls have been used to achieve the necessary properties and radiation damage resistance. Some of the key issues facing pressure vessel steels, ceramics, and fusion structural materials are described.

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).

Vanadium alloys offer many advantages including reduced neutron activation. However, activation of V–Cr–Ti alloys is controlled by impurities. This study analyzes four impurity levels for the alloy V–4Cr–4Ti. The first is an existing commercial heat of the alloy and the others are based on impurity levels attainable by three processes. One process uses a dedicated production plant; another uses a chemical recrystallization process, and the remaining one uses state-of-the-art laboratory purification. It was assumed that the alloys were irradiated as first wall and blanket structures in a fusion reactor operating for four full power years at a wall loading of 5 MW y/m2. It was found that the two purest alloys and, with special selection, the three purest alloys will meet the criteria for shallow land burial after a cooling time of 50 years. However, none of the alloys were found to meet the criteria for hands-on reprocessing after a cooling time of 100 years.

Beryllium is a material under consideration for divertor surfaces in the international thermonuclear experimental reactor (ITER). Since Be itself is not a suitable structural material for constructing the divertor, it needs to be bonded to other materials with sufficiently high thermal conductivity, which at the same time satisfy structural requirements such as adequate strength and fracture toughness. Bonding of Be to other materials is usually accompanied by (a) mismatches in thermal expansion and/or (b) metallurgical incompatibilities. In an attempt to minimize the thermal expansion mismatch we employed Fe or Ni transition layers for the bonding of Be to a copper alloy by hot isostatic pressing. To alleviate the thermodynamic incompatibility between Be and most other metals, thin Ag foil (130 μm) was used as a reaction barrier. Other experiments involved bonding Be (via an Ag reaction barrier) to a V–5Cr–5Ti (wt%) alloy. An Al–Be transition layer for bonding Be to a copper alloy was also explored. The microstructures of the interfaces were examined by optical and scanning electron microscopy. Shear tests carried out with Cu/Fe/Ag/Be and V/Ag/Be specimens indicated average room temperature shear strengths of 52 and 78 MPa, respectively. Fracture occurred usually at the Ag/Be interfaces, which were therefore the weakest link in the bonded specimens.

The U.S. structural materials program is pursuing a near-term effort related to the International Thermonuclear Experimental Reactor (ITER) and a long-term effort related to the development of reduced-activation materials for future power systems. The response of the principal ITER structural materials, austenitic stainless steels and copper alloys, to ITER irradiation conditions has been explored using fission reactor facilities. The primary radiation damage phenomena have been identified and the neutron dose and temperature conditions have been determined which will ensure adequate ductility during ITER operation. A new, reduced activation ferritic-martensitic steel has been shown to have exceptionally promising resistance to low temperature radiation embrittlement. A newly-developed V-Cr-Ti alloy has been successfully produced on a commercial scale and is being used to fabricate components for the DIII-D Radiative Divertor; recent experimental data on the irradiation performance of this alloy are discussed. Sources of radiation-induced property changes in SiC composite materials have been identified and strategics developed to improve mechanical performance of these materials during irradiation.

Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics in conjunction with Argonne National Laboratory and Oak Ridge National Laboratory has developed a plan for the utilization of vanadium alloys as part of the radiative divertor upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy. This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components and (2) demonstrating the in-service behavior of a vanadium alloy (V4Cr4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming radiative divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development efforts to support fabrication development and to resolve key issues related to environmental effects.

AbstractNew data on radiation-induced hardening, low-temperature creep and potential susceptibility (sensitization) to aqueous corrosion have been obtained on various heats of austenitic stainless steel (including type 316) irradiated at 60–400°C to 7–13 dpa. The data were obtained from spectral-tailoring reactor experiments, whose radiation-damage parameters are similar to those in the proposed International Thermonuclear Experimental Reactor (ITER) first-wall (FW) and blanket design. Austenitic stainless steels were found to increase significantly in strength at 60–330°C, to have higher irradiation-creep rates at 60°C than at 200–400°C, and to show radiation-induced changes in electrochemical properties at 200–400°C. These data on several radiation-induced property changes suggest that type 316 steel may be an adequate material for the FW of ITER. However, there is definitely a need for new data on fracture-toughness and on fatigue behavior below 400°C, as well as more data on irradiation-creep and effects of irradiation on corrosion properties, to better define temperature and dose dependencies for more detailed design analyses. Cold-working should remain an optional as-fabricated condition for the FW of ITER. Many properties of SA and CW 316 become similar after irradiation at 60–400°C. The higher initial yield-strength of CW 316 will allow higher design stress and elastic strain limits.

Primary design considerations for the Compact Ignition Tokamak toroidal field-coil cases are yield strength and toughness in the temperature range from 77 to 300 K. Type 21-6-9 stainless steel, also still known by its original Armco Steel Company trade name Nitronic 40, is the proposed alloy for this application. It has high yield strength and usually adequate base metal toughness, but weldments in thick sections have not been adequately characterized in terms of mechanical properties or hot-cracking propensity. In this study, weldability of the alloy in heavy sections and the mechanical properties of the resultant welds were investigated including tensile yield strength and Charpy V-notch toughness at 77 K and room temperature. Weldments were made in four different base metals using seven different filler metals. None of the weldments showed any indication of hot-cracking problems. All base metals, including weldment heat-affected zones, were found to have adequate strength and impact toughness at both test temperatures. Weld metals, on the other hand, except ERNiCr-3 and ENiCrFe-3 had impact toughnesses of less than 67 J at 77 K. Inconel 82 had an average weld metal impact toughness of over 135 J at 77 K, and although its strength at 77 K is less than that of type 21-6-9 base metal, at this point it is considered to be the first-choice filler metal. Phase 2 of this program will concentrate on composition refinement and process/procedure optimization for the generic ERNiCr-3 composition and will generate a design data base for base and weld metal, including tensile, fracture toughness, and crack growth rate data.