In this report a state-of-the-art study was performed to investigate the operational conditions for in-core and ex-vessel materials in a future High Performance Light Water Reactor (HPLWR) and to evaluate the potential ofexisting structural materials for application in fuel elements, core structures and out-of-core components. In the conventional parts of such a novel plant the approved materials of supercritical fossil power plants (SCFPP) can be used for the given temperature (≤ 600°C) and pressure ( 250 bar). These are either ferritic/martensitic or austenitic stainless steels. The design data for the in-core components are, however, very ambitious in comparison with conventional Light Water Reactors, especially regarding the coolant which is under high pressure (≤ 250 bar) and will have a transition from sub- to supercritical state in the core, since the water temperature increases from 290 to 510°C outlet. The expected temperature in the cladding of the fuel elements can reach up to 650°C and the neutron exposure can accumulate up to 1.13.10 2 3 n/cm 2 or 60 displacements per atom (dpa) for an envisaged target of 70 GWd/t U burnup. Taking these novel operational conditions into account an assessment of available material data was made. It is based on existing creep-rupture data, an extensive analysis of the corrosion in conventional steam power plants and the material behavior under irradiation. Compatibility between fuel and cladding materials is also considered. The potential of the different material groups available for in-core application, to be used as cladding materials of fuel elements, was further investigated by quantitative assumptions on the stress development in claddings and by a determination of the maximum achievable temperatures in dependence of cladding dimensions and the above mentioned operational conditions. More qualitative arguments on stress corrosion susceptibility are also included. It was stated that for a maximum temperature of 650°C from a standpoint of creep-rupture strength and corrosion resistance not only Ni-alloys but also austenitic stainless steels would fulfill the requirements for application as cladding materials. Taking into account specific items like the neutron absorption, the sensitivity to irradiation-induced helium embrittlement and stress corrosion cracking, it was finally concluded that the austenitic stainless steels are the better choice. The assessment has also shown that the most uncertain areas in the present analysis are the corrosion behavior under supercritical water conditions, including the effects of water chemistry/radiolysis, and the influence of a high stress state on stress corrosion and deformation mechanisms which govern the creep-rupture and creep buckling properties under irradiation. Future R&D activities should, therefore, concentrate on these open questions.

Ferritic-martensitic 9–12% CrMoVNb steels of MANET type possess a number of advantageous properties for fusion reactor application. Their optimization has led to improved creep and fracture-toughness properties. New 9–10% CrWVTa alloys have been developed by KfK/IMF in collaboration with the SAARSTAHL GmbH which have a reduced long-term activation and show in addition superior fracture toughness properties. The calculation of dose rate and other radiological parameters with the presently available FISPACT/EAF codes, extended by KfK files for sequential reactions has shown that the long-term dose-rate in these alloys is governed by the remaining ‘impurity level’ of Nb and the alloying elements W and Ta. Sequential reactions — though relevant for single alloying elements like Cr, Mn, V and N — provide only a second order effect in Fe-based alloys. A challenge for the future materials development is the production of alloys with the desired narrow specification of elements and impurities, which necessitates new ways of steelmaking.

A pressurized tube experiment was carried out to determine swelling and in-pile creep of a series of model plain Fe-15Cr-15Ni stainless steels and of different heats of the commercial German steel DIN 1.4970. Special attention was paid to the influence of minor alloying elements. Addition of Si most effectively suppresses swelling while variation of the Ti content results in a complex behaviour with the maximum suppression of swelling obtained for a Ti concentration somewhere near 0.25%. This results from a complex interaction between C and Ti. Consequently, a good swell resistance is achieved for DIN 1.4970 through a high Si content and understabilization. Good qualitative agreement has been found between the results of simulation and neutron irradiations; this does not apply to variations of major components like Cr and Ni. The intercorrelation between swelling and in-pile creep has been confirmed up to 90 dpa and can be described by an equation consisting of a SIPA term and an intercorrelation term. The discrepancy in the stress exponent of the intercorrelation term is still unsolved since the data analysis implies an exponent of 3/2, while theory and a formerly determined correlation both suggest a value of 1.

Ferritic/martensitic steels are candidate structural materials for fusion reactors because of their higher swelling resistance, higher thermal conductivity, lower thermal expansion, and better liquid-metal compatibility than austenitic steels. Irradiation effects will ultimately determine the applicability of these steels, and the effects of irradiation on microstructure and swelling, and on the tensile, fatigue, and impact properties of the ferritic/martensitic steels are discussed. Most irradiation studies have been carried out in fast reactors, where little transmutation helium forms. Helium has been shown to enhance swelling and affect tensile and fracture behavior, making helium a critical issue, since high helium concentrations will be generated in conjunction with displacement damage in a fusion reactor. These issues are reviewed to evaluate the status of ferritic/martensitic steels and to assess the research required to insure that such steels are viable candidates for fusion applications.

Tensile specimens of the tempered martensitic 12% Cr-steel MANET have been homogeneously implanted with 500 appm helium and/or 500 appm hydrogen between 80 and 500°C to investigate synergistic effects of helium, hydrogen and irradiation damage. No difference was found between He-implanted specimens and specimens simultaneously implanted with helium and hydrogen. The implantation-induced hardening decreases with increasing temperature and changes into softening above 450°C. The remaining total elongation was found to still be 2% below 400°C. However, in the temperature region of dynamic strain aging near 300°C, yield strength is close to ultimate strength, and uniform elongation drops from 3 to 0.3%. Hydrogen concentrations of 500 appm are found to modify neither strength nor ductility above 100°C. The fracture mode of hydrogen and/or helium implanted specimens remained always ductile and transgranular. The observed microstructural changes are correlated with the tensile properties.

The fusion materials programme in the EC commenced in the late seventies. To-day, a substantial part is dedicated to the preparation of a database for NET. These activities have produced a series of interesting results on structural materials for the first wall, the divertor and on protection materials in the temperature and neutron fluence range typical for NET. Other activities concern high heat flux phenomena and R&D on materials to be used in blanket test modules of NET or to be developed for long term applications as tritium breeding materials, ceramic insulators, materials for high neutron fluences and low activation materials.

The loading of the First Wall by neutrons, plasma particles and wave radiation is reviewed. For a First-Wall concept proposed by the NET-team, the cyclic thermomechanical loading conditions of austenitic (316 L) and martensitic (MANET, 1.4914) steels are compared. Due to its more favorable thermo-physical properties, the secondary stresses are lower in the martensitic steel so that its fatigue resistance should be superior. The effect of high neutron fluences has, however, yet to be determined. The ductile-brittle transition temperature of martensitic steels may limit the operational temperatures to a lower limit of about 250 ° C. Finally consideration is given to the concept of low-activation materials.

A review of mechanical properties of austenitic stainless steels is made to assess their behavior in fusion reactors. Since the first walls of fusion devices are expected to range in temperature from 100 to over 500-degrees-C, behavior over a wide range of temperatures is reviewed.For tensile properties, the neutron spectrum has little effect on strength, but appears to influence plastic deformation and ductility. The effect is apparent at temperatures above 400-degrees-C where ductility begins to increase. Ductility appears to be higher for alloys irradiated in fast reactors where little helium is produced, but even for fast reactor irradiation, ductility begins to drop above 700-degrees-C. This behavior, which is enhanced by higher helium concentrations, is believed to be a result of helium embrittlement.Irradiation creep is very weakly dependent on temperature, with nearly constant creep rates from room temperature to half the melting point. There are not yet sufficient data to determine the effect, if any, of helium on irradiation creep. However, creep rupture is exacerbated by helium since bubble formation at grain boundaries is the primary mechanism of in-reactor failure, especially at high temperatures.Fatigue is also dependent upon high temperature helium embrittlement. High temperatures, high helium concentrations, and low strain rates enhance reduction in fatigue life. At temperatures as high as 550-degrees-C, there is no apparent effect of helium even at concentrations of 500 appm at strain rates of 10(-3) s-1, but either a decrease in strain rate by a factor of 100 or an increase in helium concentration by a factor of 6 produces a degradation in fatigue life.In general, helium affects mechanical properties of austenitic stainless steels; however, it is mostly a high temperature phenomenon. Caution must be exercised at temperatures below 250-degrees-C, where uniform tensile elongation is extremely low in irradiated alloys. Low-temperature embrittlement must be investigated further to determine if incremental hardening by helium contributes.

Experimental results of a study of the effects of solute elements (carbon and silicon) on the formation of voids in nickel alloys irradiated with 100 keV nickel ions to a dose of 1 × 1016 ions/cm2 at 625°C using transmission electron microscopy are presented. Solute elements affect void formation by reducing void nucleation rate and silicon is found to be more effective than carbon. Effects of temperature in the range 475–675°C and dose (5 × 1015, 1 × 1016and 2.5 × 1016 ions/cm2) on swelling in nickel were also investigated. No voids were observed in Ni-3.6% Si at any temperature and dose. Void number density increases with increasing temperature, reaches a maximum and then decreases. With increasing dose void number density decreases due to void coalescence. Dislocation loops coexist with voids at 475°C and 525°C at a dose of 5 × 1015 ions/cm2.

The purpose of this paper is to report some observations on the basic damage processes controlling the rupture-life properties of unirradiated, irradiated, and in-reactor creep tested samples. In the first part the influence of different pre-treatments on the microstructure of 1.4970 steel after BR2 irradiation is analyzed. The effective concentration of helium near the grain boundaries is governed by the boron-bearing grain boundary precipitates. The distribution of helium bubbles within the grains is mainly influenced by dislocations rather than by secondary TiC precipitates and their interfaces. In post-irradiation creep tests, specimens showed that a considerable reduction in ductility occurs for all states of the material. In contrast to the ductility losses, the creep-rupture strengths are almost unaffected by the pre-irradiation states of the material. In the second part the published results on the in-pile creep-rupture properties of the same material are given. They indicate a marked reduction in the creep-rupture life due to irradiation. It is proposed that the in-pile creep rupture-life behaviour is caused by the SIPA creep, which operates below a specific stress level.

This paper reviews important deformation mechanisms observed in austenitic stainless steels in post-irradiation and in-reactor mechanical tests.

This paper reports on the results of development work done at Kernforschungszentrum Karlsruhe in cooperation with Interatom and ARBED-Saarstahl on materials for LMFBR fuel elements of high burn-up. The swelling resistance of the reference material for the prototype reactor SNR-300, the austenitic steel 1.4970, has been optimized by variation of minor alloying elements. An even higher swelling resistance has been found in heavy ion irradiations for a new class of austenitic steels, characterized by chromium and nickel contents of around 9 and 25 wt%, respectively. Finally, very promising results have been obtained with an optimized 12% CrMoVNb martensitic steel intended for wrapper application.