[en] Oxide dispersion strengthened (ODS) steels are notable advanced alloys with durability to a high-temperature and high-dose neutron irradiation environment because of their good swelling resistance and mechanical properties under neutron irradiation. 9–12Cr-ODS martensite steels have been developed in the Japan Atomic Energy Agency as the primary candidate material for the fast reactor fuel cladding tubes. They would also be good candidates for the fusion reactor blanket material which is exposed to high-dose neutron irradiation. In this work, modification of the manufacturing process of 11Cr-ODS steel was carried out to improve its impact property. Two types of 11Cr-ODS steels were manufactured: pre-mix and full pre-alloy ODS steels. Miniature Charpy impact tests and metallurgical observations were carried out on these steels. The impact properties of full pre-alloy ODS steels were shown to be superior to those of pre-mix ODS steels. It was demonstrated that the full pre-alloy process noticeably improved the microstructure homogeneity (i.e. reduction of inclusions and pores)

[en] Modified 316 base austenitic stainless steels (PNC316) had been developed by Japan Nuclear Cycle Development Institute as Monju core material. Irradiation creep compliance (B₀) and creep-swelling coupling coefficient (D) of PNC316 steel were previously derived from results of the pressurized tubes irradiated in FFTF/MOTA for the decision of the irradiation creep correlation. The derived values of B₀ and D were within the range B₀ and D of other austenitic steel abroad. In this study, irradiation creep strain was formulated based on point defect rate theory and calculated results were compared with the creep data of pressurized tubes irradiated in FFTF/MOTA in order to accurately evaluate irradiation creep strain for radionalization of core designing. The results can be summarized as follows. (1) The swelling-creep coupling coefficient (D) calculated from point defect rate theory decreased with swelling rate. (2) Decrease of D with swelling rate is explainable by considering not only voids but also precipitates as neutral sinks. (3) Irradiation creep compliance (B₀) might decrease gradually after swelling starts because of reduction of point defects concentration due to voids formation as neutral sinks. (4) Calculated results of SIPA creep strain were approximately consistent with PIE data of pressurized tubes if the values of point defect parameters (V_α, ΔV_α, Δμ_α) were appropriately chosen. (author)

[en] Oxide dispersion strengthened (ODS) steels are candidates for nuclear materials because of superiority in elevated temperature strength as well as high resistance to neutron-radiation damage. 9Cr-ODS ferritic-martensitic steels have been developed in JAEA for nuclear fuel cladding materials of the fast breeder reactor, which are used at high temperature and exposed to high irradiation dose. In the study, the cladding tube of 9Cr-ODS steel with the highest creep fracture strength has been developed including fabrication technology of mechanical alloying and processing ones of hot extrusion and tubing. In this summary report, technologies of nano-scale structure control of Y₂O₃ particles and meso-scale structure control of the matrix are described. The ODS steels are also candidates of structural materials for fusion reactors and advanced nuclear reactors so-called Generation IV, the developing studies for which are proceeding on a worldwide scale. (A. Hishinuma)

[en] This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be used for the fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr- and 11Cr-ODS steels have been extensively developed. Their laboratory-scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. Large-scale manufacturing technology development and mechanical testing for codification of material strength standard are on-going. As for the PNC-FMS wrapper tube, the development of a dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, codification of 316FR stainless steel and Modified 9Cr-1Mo steel is ongoing. Acquisition and collection of long-term data of base metal and welded joints are continued and evaluation methodologies are being developed to establish a technical basis for 60-year design. (author)

[en] This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be used for the fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr- and 11Cr-ODS steels have been extensively developed. Their laboratory-scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. Large-scale manufacturing technology development and mechanical testing for codification of material strength standard are on-going. As for the PNC-FMS wrapper tube, the development of a dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, codification of 316FR stainless steel and Modified 9Cr-1Mo steel is ongoing. Acquisition and collection of long-term data of base metal and welded joints are continued and evaluation methodologies are being developed to establish a technical basis for 60-year design. (author)

[en] Oxide Dispersion Strengthened (ODS) ferritic steel has excellent welling resistance and improved high-temperature strength, which are important properties for the fast reactor fuel cladding tube. Japan Nuclear Cycle Development Institute (JNC) has developed a method for microstructure control to disperse nano-sized oxide particles in equiaxed crystal grain and succeeded in producing high performance ODS ferritic steel cladding. The ring tensile and internal creep rupture strengths of the manufactured ODS ferritic claddings exhibited excellent performance far beyond that of conventional ferritic-martensitic steel (PNC-FMS) and austenitic steel (PNC316). Adequate ductility in hoop direction was also maintained. The feasibility study was also conducted for economically manufacturing process with capable of large-scale production. In order to confirm and demonstrate the ODS fuel pin integrity to high burnup conditions, the irradiation test in BOR-60 has been conducted under the framework of JNC-Russia FBR cycle cooperation and the irradiation test in JOYO is under the planning stage. (author)

[en] 14Cr-25Ni advanced austenitic steels (14Cr-25Ni steel) was developed in order to improve the swelling resistance of modified 15Cr-20Ni base austenitic steels (PNC1520). This steel is planned to be utilized as back-up material for Monju advanced core component. Seven types of cladding tubes with different amounts of alloying elements of Ti, Nb, V and P had been fabricated and the out-of-pile testing of these cladding tubes has been conducted. Among out of piles properties, tensile property and creep property of 14Cr-25Ni steels were evaluated in this study. The results can be summarized as follows. (1) Tensile properties (0.2% yield stress, ultimate tensile strength, uniform elongation, ultimate elongation) of 14Cr-25Ni steels were almost equivalent to those of PNC316 steel and PNC1520 steel in the temperature range from room temp. to 900degC. (2) The creep strength of 14Cr-25Ni steels was between PNC316 and PNC1520. It was shown that the long term creep strength at 750degC was remarkably improved by the Ti, Nb and V addition. (3) It is suspected that the improvement of creep strength of V-added-steel would be caused by stabilization of dislocation microstructure due to finely distributed carbo-nitride [Ti, Nb, V (C, N)] precipitates. (author)

[en] Irradiation behavior of ODS steel cladding tubes was evaluated for the further progress in understanding of the neutron-irradiation effects on ODS steel. Two types of ODS (9Cr–ODS_F/M, 12Cr–ODS_F) steel cladding tubes with differences in basic compositions and matrix phases were irradiated in JOYO. Post-irradiation examination data concerning hardness, ring tensile property, and microstructure were obtained. Hardness measurement after irradiation showed that there was an apparent irradiation temperature dependence on hardness for 9Cr–ODS_F/M steel whereas no distinct temperature dependence for 12Cr–ODS_F steel. Also, there was no significant change in tensile strengths after irradiation below 923 K, but those above 1023 K up to 6.6 × 10²⁶ n/m² (E > 0.1 MeV) were decreased by about 20%. TEM observations showed that the radiation-induced defect cluster formation during irradiation was suppressed because of high density sink site for defect such as initially-existed dislocation, and precipitate interfaces. In addition, oxide particles were stable up to the maximum doses of this irradiation test

[en] The dissimilar butt welding joint of 11Cr-ferritic/martensitic steel (PNC-FMS) and Type 316 austenitic stainless steel (SUS316) produced by electron beam (EB) welding was studied. This study was carried out to investigate optimization of EB welding and postweld heat treatment (PWHT) for the wrapper tube materials. Optimum EB welding conditions were a focus position of 30–40 mm and a welding speed of 1750–2000 mm/min, and optimum PWHT was performed after welding at 690°C for 60 min. As a result, no formation of δ-ferrite was observed adjacent to the fusion zone, and the mechanical properties of the welds were similar to those of the base material. In this regard, EB welding is a proper fusion welding process for dissimilar PNC-FMS and SUS316. (author)

[en] The effects of chemical compositions (titanium, oxygen) and consolidation temperature on high-temperature mechanical properties of 9Cr-oxide dispersion strengthened steel (9CrODS steel) were investigated. A possible high-temperature strengthening mechanism of 9CrODS steel was discussed based on the experimental results. Creep strength of 9CrODS steel at 973K was remarkably improved when titanium concentration was 0.35 mass%. A higher amount of added titanium than 0.2 mass% was effective for providing consistently reliable manufacturing of high strength 9CrODS steel because it reduced the effect of oxygen contamination on high-temperature strength. The fraction of elongated δ-ferrite grains, which had an ultra-fine oxide particle dispersion, tended to increase with increasing titanium. The elongated grains were considered to improve creep strength of 9CrODS steel. It was also found that creep strength was degraded by elevating the consolidation temperature from 1423 K to 1473 K. (author)