Tsuneyuki Hashimoto

Effect of Zirconium Addition to Austenitic Stainless Steels on Suppression of Radiation Induced Chromium Segregation at Grain Boundaries under Ion Irradiation

The effect of Zr addition to austenitic stainless steels on the suppression of radiation induced Cr segregation at grain boundaries under 400 keV He+ irradiation was studied. Type 316L stainless steel and steels with addition of 0.07, 0.21 or 0.41 mass% Zr were kept at 1,423K for 30 min, and then they were quenched into the water. Irradiation was done at 773K with the dose rate of 2.4×10−4dpa/s. The total dose was 0.85 or 3.4dpa. After irradiation, profiles of Cr concentration across the grain boundaries were measured using an analytical electron microscope with 1 nm beam diameter. Concentration of Cr at the grain boundary is decreased by radiation induced segregation. However, it increased with the addition of Zr, and the Cr segregation is almost completely suppressed when Zr is added more than 0.21 mass%. The effect of Zr addition on suppression of Cr segregation was analyzed focussing on the interaction between dissolved Zr atoms and point defects. The effect is based on vacancy trapping by the Zr atom...

A model for radiation-induced segregation in fcc binary alloys

Abstract A model is presented for radiation-induced segregation (RIS) in a face-centered cubic (fcc) binary alloy containing A- and B-atoms. Assuming that the interstitial in an fcc crystal takes the configuration of the 〈100〉 dumbbell, three types of interstitial dumbbells, AA-, BB- and AB-type, are considered. The present model includes the diffusion and conversion of the three types of interstitial dumbbells via an interstitialcy migration, their recombination with a vacancy, and vacancy diffusion by position exchange with a lattice atom. The fraction of AA-, BB- and AB-type dumbbells is determined through the conversions, and the AB-type mixed dumbbell plays an important role in the determination of the segregation direction. When a large number of mixed dumbbells moves toward the sinks, enrichment of the lower concentration element occurs, because the AB-dumbbell includes the same number of A- and B-atoms. When mixed dumbbell migration is rare, on the other hand, the lower concentration element can be depleted, even if the self-interstitial dumbbell composed of that element moves more quickly. Consequently, the direction of RIS depends on the mobility of point defects as well as alloy compositions. When the RIS kinetics of a Cu Au alloy is calculated with the model, good agreement is obtained with experimental results.

Re-weldability of neutron-irradiated stainless steels studied by multi-pass TIG welding

Weldability of neutron-irradiated stainless steel (SS) has been studied by multi-pass bead-on-plate and build-up tungsten inert gas (TIG) welding, simulating the repair-welding of reactor components. Specimens were submerged arc welding (SAW) joint of Type 304 SS containing 0.5 appm helium (1.8 appm in the SAW weld metal). Sound welding could be obtained by one- to three-pass welding on the plates at weld heat inputs less than 1 MJ/m in the irradiated 304 SS base metal. In the case of the build-up welding of a groove, no visible defects appeared in the specimen at a heat input as low as 0.4 MJ/m. However, build-up welding at a high heat input of 1 MJ/m was prone to weld cracking, owing to the formation of helium bubbles on grain boundaries of the base metal or dendrite boundaries of pre-existing SAW weld metal, in the area within 0.6 mm from the fusion line.

Effects of alloying elements (Mo, Si) in an austenitic stainless steel on dislocation loop nucleation under ion irradiation

Abstract Effects of alloying elements in an austenitic stainless steel on dislocation loop formation under 300 keV He+ irradiation were studied. Pure stainless steel (Fe-18 wt.% Cr-16 wt.% Ni), Mo added and Si- and/or Mo-added pure stainless steels and Type 316 stainless steel and that without Mo were used. The numbers of dislocation loops at several temperatures between 250 and 450°C were measured. For pure stainless steel, activation energy of dislocation loop nucleation is obtained as 6 × 10−20 J (0.4 eV). Mo addition does not have any effects on loop nucleation, while Si addition to the pure stainless steel promotes loop nucleation and increases the activation energy to 8 × 10−20J (0.5 eV). But when Mo is added to Fe-Cr-Ni-Si alloy, loop density and activation energy decrease to those of the pure stainless steel. These experimental results lead to the conclusion that Mo addition suppresses heterogeneous loop nucleation, in which Si acts as the nucleus, by forming Mo-Si clusters, and these ideas are used to explain differences in mechanical properties under irradiation between Types 304 and 316 stainless steels.

A theory of dislocation loop formation in binary alloys under irradiation

Abstract A model for dislocation loop formation under irradiation in a face-centered cubic (fcc) A-B alloy is presented. Besides reactions, such as defect production by irradiation, mutual recombination of an interstitial and a vacancy, dislocation loop nucleation, and their growth, conversion among the three types of interstitial dumbbells composed of A- and B-atoms is newly introduced in the formulation, considering the configurations and movements of the dumbbells. The present model is confirmed to give similar loop nucleation behavior to that derived by the conventional model in the case of a dilute alloy. Sample calculations are performed for a concentrated alloy, varying the strength of the binding between A- and B-atoms. Conversion among interstitial dumbbells is demonstrated to play an important role in the loop nucleation process. The atom that dominates loop nucleation varies with the binding of A- and B-atoms. As a result, kinetics of loop nucleation also differ.

Development of a dual-ion beam accelerator connected with a TEM for in situ observation of radiation-induced defects

A dual-ion beam accelerator connected with a TEM has been developed for in situ observation of radiation-induced defects. The system consists of a 400-kV Cockcroft-Walton accelerator, which can accelerate two different kinds of ions alternatively, and a 200-kV TEM equipped with a high-sensitivity TV camera. The ion beam from the accelerator is fed into the TEM by an electrostatic beam transport system which consists of three deflectors, two quadrupole lenses and a 57° static prism. A copper specimen is bombarded with 150-keV Ar ions. A small cascade of < 5 nm in diameter is observed for an Ar-ion current of about 85 nA/cm2. At a higher current of 1 μA/cm2, recombination, growth, overlap, and collective motions of cascades are observed during irradiation. In situ observation of argon bubbles at a grain boundary of copper gives a diameter growth rate of 2.8 × 10−2 nm/s at a dose rate of 5.3 × 1014 Ar+/cm2 s and a temperature of about 500 K.

Development of a dual ion beam system with single accelerator for materials studies

Abstract The dual ion beam accelerator system has been developed for simulation studies of neutron radiation damage of structural materials for nuclear fusion and fission reactors. One accelerator is used to accelerate two different kinds of ions, which are generated in the ion source simultaneously. One of these ions is selected alternatively by switching the magnetic field of the analyzing magnet, and is then accelerated to the desired energy value. The system is controlled by a microcomputer. The accelerator used in the system is a conventional 400 kV Cockcroft-Walton accelerator. The performance test by the acceleration of He+ and Ar+ shows that the system is capable of accelerating two ions alternatively with a switching time of less than 22 s. The beam current obtained with the microcomputer control is more than 98% of the current obtained by manual operation.

Segregation Behavior Induced by Various Particle Irradiations in Austenitic Stainless Steels

Radiation induced segregation (RIS) occurred in austenitic stainless steels was investigated to account for differences in RIS behavior under various irradiation conditions. Measured composition profiles in austenitic stainless steels irradiated by neutron in a boiling water reactor, 1 MeV electron or 60 MeV He, were analyzed by computer simulations based on the inverse-Kirkendall model. By using corresponding damage efficiency of each irradiation as an input parameter in addition to dose rate and irradiation temperature, calculations almost reproduced the measured RIS behavior. This suggested that the damage efficiency was a useful parameter of RIS for different particle irradiations. However, inconsistencies between measurements and calculations appeared in the irradiation conditions at high dose and high dose rate at about 570 K. To account for them consistently, further studies including the effect of microstructural evolution on RIS behavior were required.

Modification of the FRI Crack Growth Model Formulation from a Mathematical Viewpoint

The FRI model of crack growth, which incorporates mechanical properties into the slip oxidation mechanism of crack advance, is an extension of the well-known Ford-Andresen model. When the exponent of the oxidation current decay curve is set close to 1, however, the FRI model gives an infinite crack growth rate. Here, the oxidation decay curve integral is revised to eliminate this divergence, and modified crack growth rate equations are derived. Also presented here is a procedure for determining the oxidation current parameters from the curve-fitting to measurements of crack growth rate. Parameter value determination and crack growth calculations are illustrated for cold-worked Type 316L stainless steel.

Development of multispecimen holder for ion irradiation experiments

Abstract A specimen holder has been developed for ion irradiation experiments. Nine specimens (diameter: 3 mm, thickness: 0.1–0.3 mm) can be loaded onto the holder. The thermal resistivity between the specimen and the specimen table is reduced by setting a gold disk between them. Temperatures are controlled by a ceramic heater and cooling gas to within 1°C over the range of −50 to 550°C during ion irradiation with beam powers up to 10 W. The temperature differences between specimens are less than 5°C. The specimen assembly is insulated to allow accurate ion beam current measurements. A quartz prism is mounted in front of the specimen assembly to allow viewing of the beam shape and adjustment of the specimen position within the beamline.