Naoto Shigenaka

Effects of Hydrogen Peroxide on Intergranular Stress Corrosion Cracking of Stainless Steel in High Temperature Water, (I), Effects of Hydrogen Peroxide on Electrochemical Corrosion Potential of Stainless Steel

In order to determine effects of hydrogen peroxide on stress corrosion cracking of structural materials in the primary cooling systems of boiling water reactors, a high temperature high pressure water loop with controlled hydrogen peroxide concentrations and lower possible oxygen concentrations has been fabricated. Test specimens are installed in a stainless steel autoclave which has poly tetra-fluoro-ethylene (PTFE) inner liner to prevent decomposition of hydrogen peroxide on the autoclave surfaces. Hydrogen peroxide is injected into the autoclave inlet through the injection line which also has PTFE inner liner. The concentration of hydrogen peroxide is measured at the autoclave outlet by sampling water via the PTFE-lined sampling line. More than 65% of the injected hydrogen peroxide remains at the autoclave outlet at elevated temperature (288°C). Electrochemical corrosion potential (ECP) of stainless steel is then measured in the autoclave while changing hydrogen peroxide and oxygen concentrations. From these measurements it is concluded that, at the same oxidant concentration: (1) ECP of stainless steel exposed to hydrogen peroxide is higher than that exposed to oxygen; (2) ECP is much affected by specimen surfaces; and (3) ECP shows a hysteresis pattern for on its concentration dependence. ECP of stainless steel with an oxidized surface formed under high hydrogen peroxide concentration is much higher than that with a mechanically polished surface and it is less affected by oxidant species and their concentrations.

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...

Effects of hydrogen peroxide on intergranular stress corrosion cracking of stainless steel in high temperature water, (II) optimization of crack propagation rate measurement system

In order to determine a crack propagation rate of less than 10-8 mm/s in a 24-hour integrated measurement, major parameters of a coupled system of a constant tension specimen and crack depth measurement, based on potential drop method, have been optimized. Influences of sensor geometry, location for detecting potential drop and data processing of the ratio of signal to noise (S/N) were optimized by applying Taguchis Method. Then a suitable sensor geometry and data processing method were proposed to get a robust measurement system with higher sensitivity and lower susceptibility for geometrical and procedural fluctuations. By applying the optimal crack propagation rate measurement system, it was confirmed that a crack propagation rate of lxlO-8 mm/s can be measured under a low concentration condition of hydrogen peroxide with less than a 20% error by a 24-hour integrated measurement.

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.

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.

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.

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.

Effects of gamma-ray irradiation on crevice corrosion repassivation potential of stainless steel in high temperature diluted simulated seawater

The crevice corrosion repassivation potentials (ER,CREV) of type 304 stainless steel (304 SS) were measured in high temperature (373–553 K), diluted simulated seawater under gamma-ray irradiation, in order to confirm the effects of gamma-ray irradiation on the crevice corrosion behavior of a representative stainless steel in seawater. Overall, for high temperatures, the ER,CREV values decreased with increasing chloride ion concentration, which was the same as the behavior observed under the non-irradiated condition. The ER,CREV values measured under gamma-ray irradiation were the same or slightly higher than ER,CREV values measured under the non-irradiated condition when the [Cl−] was the same. Consequently, it was confirmed that the threshold potential of crevice corrosion of 304 SS for the gamma-ray irradiation of 1.8 kGy at least did not deteriorate compared with the non-irradiated condition. Under the conditions of this work (seawater composition, [Cl−] range, dose rate, absorbed dose, flow rate, etc.), the crevice corrosion of 304 SS could be suppressed by maintaining the potential below the threshold potential which was determined approximately as −0.3 V vs. SHE even for the irradiated condition at temperatures up to 553 K.