Kyoichi Asano

Post-irradiation annealing effects on microstructure and helium bubbles in neutron irradiated type 304 stainless steel

Abstract Helium bubble nucleation and growth processes were studied at elevated temperatures on type 304 stainless steel which had been neutron irradiated to 1.4 × 1026 n / m 2 ( E > 1 MeV) and annealed at 400, 550, 650 and 900°C for 1 h. After annealing, specimens for microstructural observation were prepared and observed (by transmission electron microscope (TEM).) Radiation defects were present at high density in the as-irradiated specimens and annealed out with increasing annealing temperature. Helium bubbles became visible where defects were annealed out (above 650°C). The bubbles grew preferentially at dislocations and grain boundaries. At the grain boundaries, most of the helium bubbles formed on grain boundary dislocations. It was concluded that during post-irradiation annealing, the dislocations which remained after annealing played an important role in helium bubble growth, even in the grain boundaries.

Effects of hydrogen on mechanical properties in irradiated austenitic stainless steels

Abstract To investigate the hydrogen effect on mechanical properties of solution annealed Type 304 stainless steel, tensile tests of neutron irradiated materials were conducted after a hydrogen charging and discharging process (hydrogen treatment). Elongation was less with increasing neutron fluence after hydrogen treatment than that of as-irradiated specimens. Intergranular cracking occurred by the hydrogen treatment in heavier irradiated specimens, in which the Cr depleted zone along grain boundary was observed. Embrittlement and intergranular cracking after the hydrogen treatment were estimated to be attributed to the Cr depleted zone at the grain boundary due to neutron irradiation.

Stress corrosion cracking and intergranular corrosion of austenitic stainless steels irradiated at 323 K

The effect of neutron irradiation on intergranular corrosion (IGC) and stress corrosion cracking (SCC) susceptibility was investigated in several austenitic stainless steels irradiated up to 3.0 × 1025 n/m2 (E > 1 MeV) at about 323 K. The IGC of commerical purity (CP) 304 and CP304L in HNO3/Cr6+ solution was enhanced by neutron irradiation, while HP304 was immune to corrosion. The stainless steels except nuclear grade 316 (316NG) failed intergranularly after neutron irradiation by SSRT. HP304 showed higher IASCC susceptibility than CP304 and CP304L. The occurrence of IASCC in both HP and CP 304 indicated that radiation enrichment of impurities might not be main contributor to IASCC. 316NG had better SCC resistance than 304 stainless steels under the present irradiation conditions. IASCC susceptibility could be correlated to the austenite phase stability.

Hydrogen analysis and slow strain rate test in Ar gas for irradiated austenitic stainless steel

Abstract Hydrogen concentration in austenitic stainless steel irradiated with neutrons in boiling water reactors (BWRs) was measured and the effect of hydrogen in the austenitic stainless steel on intergranular cracking was investigated by the slow strain rate test (SSRT) in Ar gas. The hydrogen concentration decreased at low neutron fluences and increased at high neutron fluences. The decrease was attributed to the effect of heating or γ-ray irradiation at the early stage of reactor operation. The increase at high fluences was considered mainly due to the generation of hydrogen by nuclear transmutation. Intergranular cracking was not found for the specimen irradiated to a high fluence ( 1.4×10 26 n / m 2 ) in the SSRT at a very low strain rate ( 1.0×10 −8 s −1 ). This meant that the hydrogen concentration was too small to induce cracking, or hydrogen could not diffuse because of being trapped in irradiation defects at the test temperature.

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.

Investigation of Effect of Pre-Strain on Very High-Cycle Fatigue Strength of Austenitic Stainless Steels

The austenitic stainless steels have excellent mechanical and chemical characteristics and these materials are widely used for the main structural components in the nuclear power plants. A part of structural components using these materials is considered to have strain-history by machining, welding and etc in the process of manufacturing and these parts would be hardened because these materials have a remarkable work-hardening property. On the other hand, conventional studies for the fatigue strength used to be investigated by the results of fatigue tests applying normal specimens without the effect of hardening by pre-strain. This paper describes the effect of large pre-strain on very high cycle fatigue strength of the materials in consideration for the evaluation of strength of actual structures in the nuclear power plants. In order to achieve this purpose, the fatigue tests were carried out with strain hardened specimens. The material served in this study was type SUS316NG. Up to ±20% pre-strain was introduced to the round bar shaped materials by tension and compression load test, and the materials were mechanically machined to the hourglass shaped smooth specimens. On the other hand, the pre-strain of some specimens were introduced after machining so as to study the influence of roughness of the surface of the specimens for the fatigue property. Fatigue tests were conducted by ultrasonic and rotating-bending fatigue test machines and conditions were decided by preliminary examinations to control temperature elevation of the specimen during the fatigue test. The S-N curves obtained from fatigue tests show that increase in magnitude of the pre-strain cause increase in the fatigue strength of the material and this relationship is independent of type of the pre-strains of tension and compression. Though all specimens were fractured by the surface initiated fatigue crack, only one specimen was fractured by the internal crack and so-called “fish-eye” was observed on the fracture surface. However, the internal fracture of the SUS316NG does not cause sudden drop of the fatigue strength. Also, the Vickers hardness tests were carried out to discuss the relationship between fatigue strength and hardness of the pre-strained materials. It is found that the increase in fatigue limit of the pre-strained materials strongly depend on the hardness derived from the indentation size equals to the scale of stage I fatigue crack.Copyright