Ryuta Kasada

High Burnup Fuel Cladding Materials R&D for Advanced Nuclear Systems: Nano-sized oxide dispersion strengthening steels

Cladding materials development is crucial to realize highly efficient and high-burnup operation over 100GWd/t of so called Generation IV nuclear energy systems, such as supercritical-water-cooled reactor (SCWR) and lead-cooled fast reactor (LFR). Oxide dispersion strengthening (ODS) ferritic/martensitic steels, which contain 9–12%Cr, show rather high resistance to neutron irradiation embrittlement and high strength at elevated temperatures. However, their corrosion resistance is not good enough in SCW and in lead at high temperatures. In order to improve corrosion resistance of the ODS steels in such environment, high-Cr ODS steels have been developed at Kyoto University. An increase in Cr content resulted in a drastic improvement of corrosion resistance in SCW and in lead, while it was expected to cause an enhancement of aging embrittlement as well as irradiation embrittlement. Anisotropy in tensile properties is another issue. In order to overwhelm these issues, surveillance tests of the material performance have been performed for high Cr-ODS steels produced by new processing technologies. It is demonstrated that high-Cr ODS steels have a high potential as fuel cladding materials for SCWR and LFR with high efficiency and high burnup.

High resistance to helium embrittlement in reduced activation martensitic steels

Abstract Reduced activation martensitic steels (RAMSs) are the prime candidate structural material for the DEMO reactor and beyond where the material has been considered to suffer severe embrittlement caused by high-dose neutron irradiation and several thousands appm of transmutated helium. However, recent several works show high resistance to helium embrittlement of RAMS. Good performance of RAMS in the presence of rather high concentration of helium is considered to be due to high trapping capacity for helium atoms in the martensitic structure that consists of dislocations, lath boundaries, grain boundaries and carbide/matrix interfaces. To make clear the role of dislocations in trapping helium, thermal helium desorption spectra were measured for iron specimens annealed at different temperatures after cold work. A desorption peak, which increased its height with increasing dislocation density, was observed at around 550 °C, suggesting that dislocations trap helium atoms. A molecular dynamics simulation study for investigating the helium trapping behavior at helium–vacancy complexes suggests that helium is rather strongly bound to the complexes and increases the binding energy of vacancy to the complex, resulting in increasing stability of the complexes at elevated temperatures by reducing thermal emission of vacancies.

Annealing behavior of irradiation hardening and microstructure in helium-implanted reduced activation martensitic steel

Abstract Post-implantation annealing behavior was investigated for a reduced activation martensitic steel (RAMS), which was homogeneously implanted with 580 at.ppm of helium by cyclotron utilizing energy degrader at below 429 K. Post-implantation isochronal annealing caused no age hardening but the gradual recovery of the hardening even above 673 K, while the neutron irradiated specimen showed a complete recovery of the hardening by the annealing above 673 K. Two-component analysis of positron lifetime measurements along with hardness measurements indicated that long lifetime component ( τ 2 ) in the helium-implanted steel still existed after annealing up to 873 K. Evolution of helium bubbles during annealing was examined by TEM and it was revealed that helium bubbles tend to be formed at lath boundaries by the annealing above 723 K. Helium desorption was observed by the annealing above 773 K where recovery of the hardening began.

Recent progress of tungsten R&D for fusion application in Japan

The status of ongoing research projects of tungsten R&D in Japan is summarized in this paper. For tungsten material development, a new improved fabrication technique, the so-called superplasticity-based microstructural modification, is described. This technique successfully improved fracture strength and ductility at room temperature. Recent results on vacuum plasma spray W coating and W brazing on ferritic steels and vanadium alloys are explained. Feasibility of these techniques for the manufacture of the blanket is successfully demonstrated. The latest findings on the effect of neutron damage in tungsten on T retention and on the change in mechanical and electrical properties are described. Retention characteristics for neutron-damaged W were different compared to those for ion-damaged W. Upon neutron irradiation, tungsten alloys containing transmutation elements of W (Re and Os) show changes in properties that are different compared with those shown by pure W. The effects of mixed plasma exposure (D/He/C) are described. Both D/He and D/C mixed ion irradiations significantly affect ion-driven permeation in W. He bubble dynamics play a key role in nano-structure formation on the W surface.

Enhancement of irradiation hardening by nickel addition in the reduced-activation 9Cr–2W martensitic steel

Reduced-activation martensitic (RAM) steels with and without an addition of 1% Ni were irradiated in a so called multisection-multidivision controlled irradiation capsule in the JMTR at 220°C up to 0.15 dpa. The 1/4 power dependence of the irradiation hardening on neutron dose was observed for the specimens irradiated in the controlled capsule. A part of the specimens were simultaneously irradiated in the capsule out of the reactor core where the irradiation temperature was considered to be lower than 170°C. The out of-reactor core irradiation induced a tremendous irradiation hardening as much as 350 MPa in the Ni added RAM steel but only 120 MPa of the hardening in the unadded RAM steel. The tremendous irradiation hardening was never observed following the irradiation at 220°C. As for the results of positron annihilation measurements, no significant effect of the Ni addition was observed in the life time spectrum. Post-irradiation annealing studies indicate that the irradiation hardening observed in the Ni added RAM steel begins to recover at 190°C and diminishes after the annealing at 250°C.

Effect of helium implantation on mechanical properties and microstructure evolution of reduced-activation 9Cr–2W martensitic steel

Abstract A reduced-activation martensitic steel was implanted with helium up to 580 at. ppm by using 36 MeV α-beam between 353 and 423 K along with displacement damage up to 0.226 dpa. The implantation-induced increase in ductile–brittle transition temperature (DBTT) was estimated to be 98 K for the standard charpy V-notched (CVN) specimen implanted with 580 at. ppm He, through the conversion of small punch (SP) test results by an empirical relationship. It is clarified from comparison with neutron irradiation data that the increase in DBTT as well as implantation-induced hardening is interpreted simply in terms of displacement damage, suggesting that there is no significant effect of helium on both the irradiation hardening and the fracture toughness of the steel. No fracture mode change by the helium implantation was observed in the SP tests, showing a complete cleavage fracture mode in the lower shelf energy region.

Theromophysical Properties and Microstructure of Plasma-Sprayed Tungsten Coating on Low Activation Materials

Abstract Tungsten (W) coating on various low activation materials, such as ferritic steel (F82H), oxide dispersion strengthened (ODS) steel, and vanadium alloy NIFS-HEAT-2 (NH2) was successfully demonstrated by the vacuum plasma spray (VPS) process. Void and crack-type defects were observed in VPS-W. The mass density of VPS-W at room temperature (RT) was ∼90 % of the bulk W (sintered W). The thermal diffusivity and thermal conductivity of VPS-W from RT to 800 °C were 30∼50 % of the bulk W, while the linear expansion coefficient and specific heat of VPS-W were similar to the bulk W. The thermal conductivity of VPS-W was significantly lower than the bulk W, but was still larger than the NH2 substrate. There was no significant thermal contact resistance at the interface between W coating and NH2 substrate. Thus, the heat transfer properties of NH2 will not be degraded by the W coating with the VPS process.

Grain Boundary Phosphorus Segregation in Thermally Aged Low Alloy Steels

Intergranular embrittlement due to grain boundary segregation of phosphorus is recognized as one of the potential degradation factors in irradiated reactor low alloy steels at high neutron fluence. In this study, low alloy steels thermally aged at 400-500°C were investigated to evaluate the correlation between phosphorus segregation and intergranular embrittlement. Phosphorus segregation determined using Auger electron spectroscopy increased after thermal aging above 450°C and was in good agreement with the calculated value based on McLeans model. No influence of thermal aging on tensile properties or hardness was observed. The ductile brittle transition temperature determined using a one-third size Charpy impact test increased at a P/Fe peak ratio of 0.14. These results indicated that there is a threshold level of phosphorus segregation for non-hardening embrittlement. The ductile to brittle transition temperature (DBTT) increased with the proportion of intergranular fracture, so this result shows that there is a relationship between DBTT and the proportion of intergranular fracture. The fracture stress decreases due to non-hardening embrittlement on the thermally aged material with high proportion of intergranular fracture.

Assessment of Neutron Irradiation-Induced Grain Boundary Embrittlement by Phosphorous Segregation in a Reactor Pressure Vessel Steel

The materials used were two sorts of reactor pressure vessel steels (RPVSs), which contain different amounts of impurity phosphorous (P) and copper (Cu). The specimens for Charpy V-notch impact tests and Auger electron spectroscopy (AES) were irradiated in the Japan Materials Test Reactor (JMTR) at 290°C up to fluences of 6 × 1021 and 1 × 1024 n/m2 using a multi-division temperature control capsule which enables removal of a part of the sub-capsules during operation of the reactor. Neutron irradiation resulted in a significant shift in the ductile-brittle transition temperature (DBTT) accompanied by a large irradiation hardening in the high P and high Cu steel. The AES measurements following the irradiations revealed that almost no phosphorous segregation occurred at grain boundaries. The DBTT shifts by neutron irradiation were reasonably interpreted in terms of a so-called hardening mechanism. The SEM observations of the fractured surface indicated that a very small amount of grain boundary fracture was induced at the irradiation conditions, and resultantly no grain boundary embrittlement was observed. Based on the results of irradiation experiments as well as long-term thermal aging experiments beyond 16 000 h, neutron irradiation-induced grain boundary embrittlement is considered to rarely happen to the RPVS that contains 110 wppm of phosphorous.

Modeling of microstructure evolution and mechanical property change of reduced-activation martensitic steel during varying-temperature irradiation

The effects of varying-temperature irradiation on mechanical properties and microstructure of reduced-activation 9Cr martensitic steels (RAMS) were investigated by means of micro-Vickers hardness tests and positron annihilation lifetime spectrometry. In case of stepwise increasing-temperature irradiation of 473/623 K, irradiation hardening accumulated at the lower temperature still existed after the elevation in temperature, while in case of the 493/693 K varying irradiation, the low temperature irradiation hardening disappeared after the elevation in irradiation temperature. In both the varying-temperature irradiations, it was observed that microvoids disappeared after the elevation in irradiation temperature. Result of a computer simulation of the evolution of defect clusters using rate theory were in good agreement with the experimental results when I-clusters were considered to be a factor controlling irradiation hardening.