Hideo Kayano

Microstructure and impact properties of ultra-fine grained tungsten alloys dispersed with TiC

Abstract In order to improve both the low temperature toughness and the resistance to embrittlement by recrystallization and irradiation in currently available tungsten and its alloys, ultra-fine grained tungsten alloys with TiC additions of 0.2 and 0.5 wt% were developed by mechanical alloying and hot isostatic pressing. It is shown that the impact toughness of the developed alloys is very sensitive to the magnitude of relative density and is greatly improved by increasing its value. An alloy with 0.2 wt% TiC, which has the highest relative density of 99.5% among the developed alloys, exhibits a much lower ductile-to-brittle transition temperature and higher strength than pure tungsten which has a relative density of 100%. For the alloy, recrystallization and grain growth occur during 1-h heating between 2273 and 2473 K, much higher than the equivalent temperature range for pure tungsten. Increasing the TiC content to 0.5 wt% makes the alloy more resistant to recrystallization and grain growth.

R&D of low activation ferritic steels for fusion in japanese universities

Abstract Following the brief review of the R&D of low activation ferritic steels in Japanese universities, the status of 9Cr-2W type ferritic steels development is presented. The main emphasis is on mechanical property changes by fast neutron irradiation in FFTF. Bend test, tensile test, CVN test and in-reactor creep results are provided including some data about low activation ferritic steels with Cr variation from 2.25 to 12%. The 9Cr-2W ferritic steel, denoted as JLF-1, showed excellent mechanical properties under fast neutron irradiation as high as 60 dpa. As potential materials for DEMO and beyond, innovative oxide dispersion strengthened (ODS) quasi-amorphous low activation ferritic steels are introduced. The baseline properties, microstructural evolution under ion irradiation and the recent progress of new processes are provided.

The inhomogeneous deformation behaviour of neutron irradiated Zircaloy-2

Annealed Zircaloy-2 specimens irradiated to 3.2 × 1019 n/cm2 (E > 1 MeV) at ≈425 K were tensile-tested, together with unirradiated material, between 298 and 673 K in vacuum. The surface and microstructure of deformed specimens were observed using projector, optical and transmission electron microscope. Metallographie examinations or irradiated samples showed that localized deformation bands occur at intervals during deformation to the ultimate tensile stress between 473 and 623 K, while in the room temperature deformation, the inhomogeneity in metallographic features was characterized by microscopic dislocation channeling structure, without showing localized band on the surface of deformed specimen. From the shape of the stress-strain curves and metallographic features, it was concluded that the first localized band occurred after a small amount of strain, resulting in yielding and that the flow stress increased monotonically with further yielding. Evidences from temperature dependent mechanical properties of irradiated and unirradiated samples indicate that the radiation-anneal hardening phenomena are of significance in a range of temperature at which localized bands are formed, particularly pronounced at 553–593 K.

Development of Mo alloys with improved resistance to embrittlement by recrystallization and irradiation

Abstract In order to overcome the recrystallization embrittlement and irradiation embrittlement in Mo and W, which are major problems for their fusion applications, the basic idea of alloy design and microstructure control was presented. By applying the idea to Mo, ultra-fine grained Mo alloys with very fine TiC particles mostly existing at grain boundaries were developed for TiC additions of 0.1 to 1.0 wt%. Impact three-point bending tests showed that before and after recrystallization and fast-neutron irradiations to 0.08 dpa the developed alloys exhibit much lower ductile-brittle transition temperature (DBTT) and higher strength than TZM. The developed alloys also showed much higher resistance to recrystallization and grain growth. Both resistance increased with increasing TiC content; in the irradiated state the 1.0 wt% TiC-added alloy showed a DBTT lower by more than 200 K than TZM. The cause of the observed improvement was discussed.

Radiation induced conductivity of ceramic insulators measured in a fission reactor

In-reactor measurements of the long-term change in the electrical conductivity of α-alumina were carried out in the JMTR fission reactor. Special attention was focussed on the effect of applied voltage on electrical degradation. Two experiments were carried out for 96 and 48 reactor full power days at 600–630 K and 770–800 K, with an applied electric field of 500 V/m AC and 500 kV/m DC, respectively. A long-term increase in electrical conductivity was observed, which is thought to be radiation induced electrical degradation, RIED.

Effect of specimen size on the ductile-brittle transition behavior and the fracture sequence of 9Cr-W steels

Abstract The effect of specimen size on the ductile-brittle transition behavior and the fracture sequence were investigated by means of Charpy absorbed energy measurement and fractography, using the full size, the half size and the one-third size V-notch specimens of 9Cr-W steels. The steels used are reduced-activation ferritic steels for fusion reactor structures. Attempts were made to correlate the impact data between the different specimen sizes by using normalizing parameters, such as nominal fracture area and nominal fracture volume for the upper shelf energy and ligament size for the ductile-brittle transition temperature. Fractography showed a similar fracture sequence for the three different sizes of the specimens.

Effect of neutron irradiation on low temperature toughness of TiC-dispersed molybdenum alloys

Abstract The effect of neutron irradiation on the low temperature toughness of TiC-dispersed molybdenum alloys that were recently developed for fusion applications was studied. Miniaturized bend bar specimens, 1 by 1 by 20 mm, of both the alloys with TiC additions to 1.0 wt% and TZM were irradiated to 8 × 10 23 n/m 2 ( E n > 1 MeV) at controlled cyclic temperatures between 573 and 773 K in the JMTR and were tested by impact three-point bending. It was shown that under irradiation conditions the developed alloys exhibited much higher toughness than TZM and the toughness increased with increasing TiC content. In addition, the ductility of the alloy with 1.0 wt% TiC increased drastically by the irradiation, by about five times, regardless of the occurrence of significant irradiation hardening. This notable ductilization bX irradiation is the first found in materials and is discussed in connection with radiation enhanced precipitates.

Irradiation hardening of Fe-Cr alloys

Abstract The mechanism of irradiation hardening of Fe and Fe-Cr simple α-phase alloys was studied by means of tensile tests. Some commercial ferritic and martensitic steels were also examined. The irradiation hardening of Fe was explained by the presence of small dislocation loops formed at the sites of cascade damage collapse, while that of Fe-Cr alloys was thought to be due to Cr-rich precipitates formed at the same sites. This hardening mechanism suggests that there is a linear increase with Cr content in the athermal component of the irradiation hardening of Fe-Cr alloys. This mechanism will also contribute to the irradiation hardening of Fe-Cr ferritic and martensitic steels.

Fission reactor irradiation of materials with improved control of neutron flux-temperature history

Abstract Eliminating the deficiency in the conventional temperature control, irradiation of materials with the Japan Material Testing Reactor is performed with an newly-designed in-core irradiation ring with which the sample temperature can be maintained regardless of the reactor power. The defect microstructures in various materials are compared with those introduced by irradiation with conventional control. The strong influence of the transient lower temperature irradiation during the start-up of the reactor is obvious for the samples irradiated with conventional control. The influence is fully understood from the temperature dependence of the microstructure evolution mechanism. The necessity of improved control is reconfirmed.

Mechanical properties changes of Fe-Cr alloys by fast neutron irradiation

Abstract The changes in the mechanical properties of Fe-Cr ferritic, single α-phase alloys, containing 0–15 wt% Cr, by fast neutron irradiation were studied by means of tensile tests in the temperature range between 77 and 283 K, mainly with regard to the effects of the Cr content. The temperature dependence of the yield stress decreased with increasing Cr content. The irradiation raised the thermal activated component of the yield stress in low Cr alloys severely, but not in high Cr alloys. The fracture stress in the brittle fracture mode increased with Cr content up to 10 wt%, beyond which it decreased. The frequency of twinning deformation showed an opposite tendency to the fracture stress with or without irradiation. The transition temperature of the unstable plastic flow, which occurred intensively after irradiation, decreased with increasing Cr content. The last condition mainly governed the irradiation embrittlement of Fe-Cr alloys determined from absorbed energy-temperature curves in tensile testing. Thus, high Cr alloying can make the steel more resistant to irradiation embrittlement.