K. Shiba

Larson–Miller Constant of Heat-Resistant Steel

Long-term rupture data for 79 types of heat-resistant steels including carbon steel, low-alloy steel, high-alloy steel, austenitic stainless steel, and superalloy were analyzed, and a constant for the Larson–Miller (LM) parameter was obtained in the current study for each material. The calculated LM constant, C, is approximately 20 for heat-resistant steels and alloys except for high-alloy martensitic steels with high creep resistance, for which

Irradiation effects on reduced activation ferritic/martensitic steels?tensile, impact, fatigue properties and modelling

C \approx 30

Ni-Doped F82H to Investigate He Effects in HFIR Irradiation

. The apparent activation energy was also calculated, and the LM constant was found to be proportional to the apparent activation energy with a high correlation coefficient, which suggests that the LM constant is a material constant possessing intrinsic physical meaning. The contribution of the entropy change to the LM constant is not small, especially for several martensitic steels with large values of C. Deformation of such martensitic steels should accompany a large entropy change of 10 times the gas constant at least, besides the entropy change due to self-diffusion.

Status and key issues of reduced activation ferritic/martensitic steels as the structural material for a DEMO blanket

Abstract A toughness-improved type of F82H steel called F82H mod3 has been developed, and the material properties and irradiation behavior have been examined. The significant modification of the chemical composition is the reduction of Ti (<10 ppm) and N (<20 ppm) as impurities and the increase of Ta (0.1%) as an alloying element. The ductile-to-brittle transition temperature (DBTT) is improved to -90°C from -45°C for F82H IEA without change in strength. However, the creep rupture time of F82H mod3 was 1/10 of F82H IEA. Another feature of the F82H mod3 is the stability of the material properties. Higher temperature normalization (1080°C) degrades the DBTT only to -80°C due to grain coarsening without large change in strength. It is quite important for large-scale production of the material in high quality. Preliminary neutron irradiation experiments up to 17 dpa showed better irradiation resistance to changes in fracture toughness than F82H IEA.

Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

At temperatures below 400 °C, irradiation often causes hardening and reduction of elongation as well as toughness degradation to a considerable degree. Data, however, indicate that these changes remain in manageable ranges for ITER-TBM application. Moreover, the saturation tendency of these changes with neutron dose suggests that some of the reduced activation ferritic/martensitic steels are feasible even for future DEMO applications. It is also stressed that the development of a design methodology that is compatible with the large irradiation induced property changes is essential to enable these applications. Modelling activities for the macroscopic mechanical response are expected to play key roles in design methodology development. Macroscopic models of plasticity (a constitutive equation) and cyclic softening behaviour after irradiation are discussed. The significance of the models for estimating microstructural change during irradiation and beneficial effects of the heat treatment for irradiation performance are also introduced.

Joining technologies of reduced activation ferritic/martensitic steel for blanket fabrication

An overview of the present status of development of fusion nuclear technologies at Japan Atomic Energy Research Institute is presented. A tritium handling system for the ITER was designed, and the technology for each component of this system was demonstrated successfully. An ultraviolet laser with a wavelength of 193 nm was found quite effective for removing tritium from in-vessel components of D-T fusion reactors. Blanket technologies have been developed for the test blanket module of the ITER and for advanced blankets for DEMO reactors. This blanket is composed of ceramic Li2TiO3 breeder pebbles and neutron multiplier beryllium pebbles, whose diameter ranges from 0.2 to 2 mm, contained in a box structure made of a reduced-activation ferritic steel, F82H. Mechanical properties of F82H under a thermal neutron irradiation at up to 50 displacements per atom (dpa) were obtained in a temperature range from 200 to 500°C. Design of the International Fusion Materials Irradiation Facility (IFMIF) has been developed to obtain engineering data for candidate materials for DEMO reactors under a simulated fusion neutron irradiation up to 100 to 200 dpa, and basic development of the key technologies to construct the IFMIF is now under way as an International Energy Agency international collaboration.

Technical issues related to the development of reduced-activation ferritic/martensitic steels as structural materials for a fusion blanket system

ABSTRACT Nickel-doped F82H alloys have been fabricated to simulate He production due to fusion neutrons in fission reactor irradiation. 1.2Ni and 1.4Ni alloys were tempered at 750°C without re-austenitisation. Expected He production in 1.4% Ni alloy irradiated in HFIR target position is about 400 appm at 40 dpa. Results of tensile and Charpy impact tests of these alloys show that their mechanical properties are similar to those of original F82H, although 0.2% proof stresses of Ni-doped alloys were 50 Mpa smaller than that of F82H. Small amount of two isotope tailored alloys including 1.4wt% Ni are also prepared using 58Ni and 60Ni. Chemical analyses and Charpy impact tests of the mock-up heat suggest that the fabrication of these small heats was successful.