The main objective of this paper is to summarize CEA data recently obtained on different kinds of steels irradiated at 325°C in the Osiris experimental reactor in a typical PWR environment, that is, pressurized water -- P=155 bars -- with controlled PWR type chemistry. The different steels studied can be classified in four groups: 304/316 type austenitic stainless steels, conventional 9-12%Cr(Mo,V,Nb) and reduced activation 7.5-11%Cr(W,V,Ta) martensitic steels, and two ferritic-martensitic alloys strengthened by oxide dispersion (ODS). Some of those steels have been included with different initial metallurgical conditions, that is, (1) for austenitic steels : solution annealed and cold-worked structures; (2) for martensitic steels : tempered, cold-worked and as-quenched martensitic structures. This experimental irradiation, named "Alexandre", has been carried out in the Osiris experimental reactor (under a mixed fast/thermal neutron flux) for different neutron fluence levels with a maximum irradiation damage of ~9dpa. The main results obtained are discussed, with a special emphasis on the chemical composition and initial metallurgical condition effects on the tensile and uniform corrosion properties of both conventional and reduced activation chromium-rich martensitic steels.

An experimental irradiation, named "Alexandre," has been carried out in the Osiris experimental reactor to perform a generic study on the mechanical behavior after irradiation at 325°C of different kinds of steels suitable for use as irradiated components in a nuclear reactor [1]. The irradiated steels were austenitic stainless, martensitic (conventional and reduced activation), and ferritic-martensitic Oxide Dispersion Strengthened steels in various initial metallurgical conditions. The final dose was 9 dpa, which represents nearly a "saturation" dose for the hardening/embrittlement of both austenitic and martensitic steels. At this dose, the Yield Strength and the Ultimate Tensile Strengths are almost equal, and strong localization of the plastic deformation is often observed.

A dedicated series of irradiation experiments has been executed in the high flux reactor (HFR) in Petten (Netherlands) to evaluate the irradiation behavior of ferritic-martensitic 9-12%Cr steels at temperatures in the range of 70°C to 370°C and damage dose levels up to 3 dpa. Materials investigated in the program comprise Mod.9%Cr (9Cr-1Mo-0.2V-0.08Nb), HT9 (12Cr-1Mo-0.5W-0.5Ni-0.3V), MANET type steel (10Cr-0.5Mo-0.6Ni-0.2V-0.15Nb), NF616 (9Cr-2W-0.5Mo-0.2V-0.07Nb) and HCM12A (12Cr-2W-1Cu-0.4Mo-0.3Ni-0.2V-0.05Nb). The 9-12%Cr steels show severe hardening and ductility reduction at room temperature (RT) after neutron irradiation. Strength, ductility and toughness of material irradiated at 70°C gradually recover with increase in test temperature. Similar ductility trends of material irradiated at 300°C are observed as for unirradiated material, but recovery with increase in temperature is not observed below temperatures of 400°C. The 9%Cr steels show less hardening and reduction in ductility than the 10-12%Cr steels. The reduction of area and 0.2% yield stress correlate well with the upper-shelf energy and ductile-to-brittle transition temperature, respectively, both for the unirradiated and irradiated condition. In general, the 9%Cr steels show more resistance to irradiation at 300°C than the 10-12%Cr steels with respect to fracture toughness.