Koji Fukuya

Corrosion Behavior of Stainless Steels in Simulated PWR Primary Water : Effect of Chromium Content in Alloys and Dissolved Hydrogen

The structure and composition of surface oxide films on austenitic stainless steels in hydrogenated high-temperature water were examined by changing the chromium content in alloys and the concentration of dissolved hydrogen in high-temperature water. Auger electron spectroscopy, X-ray diffraction and analytical transmission electron microscopy revealed that the oxide films had a double-layer structure: ironbased spinels as the outer layer and chromium-rich spinel oxide as the inner layer. Increasing the chromium content suppressed the corrosion rate and produced fine oxide particles with a higher chromium concentration in the inner layer. Increasing the concentration of dissolved hydrogen enhanced the corrosion rate without a notable change in oxide structure. These influences are considered to originate from changes in cation diffusion through the inner layer, such as a decrease in the lattice diffusion of iron in the inner layer due to a higher concentration of chromium in the oxide as a diffusion barrier for a high chromium content in the alloys and due to a lower oxygen partial pressure for a higher concentration of dissolved hydrogen.

Current understanding of radiation-induced degradation in light water reactor structural materials

Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.

Evolution of Microstructure and Microchemistry in Cold-worked 316 Stainless Steels under PWR Irradiation

The evolution of microstructures and microchemistry was examined by transmission electron microscopy in cold-worked SUS 316 stainless steel components irradiated in a pressurized water reactor to 1–73 dpa at 565–596 K. Homogenous nucleation of dislocation loops, helium bubbles and γ′ precipitates was detected. The dislocation loops consisted of a high density of Frank loops and black dots. The black dots are considered to be small Frank loops, some fraction of which could be vacancy-type. The size distribution showed a double peak, which remained unchanged up to 73 dpa, suggesting that the dislocation structure was saturated under a balance of nucleation and disappearance. The helium bubbles were extremely dense and fine, resulting in swelling of less than 0.1%. Such bubble structure was formed under irradiation with a high He/dpa ratio. The γ′ precipitation was detected at doses higher than 4 dpa with an increasing density for higher doses. The measured radiation hardening was almost explained by these visible microstructural features. Radiation-induced segregation at grain boundaries was confirmed to continuously develop up to 73 dpa whereas no significant segregation could be detected at Frank loops.

In Situ Studies of the Effects of Ion Beams on Materials Using the Electron Microscope Ion Beam Interface

To investigate microstructural evolution by cascade damage produced by energetic heavy particles, we have built a facility which is capable of observing damaged structure introduced by heavy ions in situ in an electron microscope. A 400 kV Cockcroft-Walton type heavy ion accelerator with a Danfysik 911A heavy-ion source has been combined with a 200 kV electron microscope. Heavy ion beams of energies up to 400 keV can bombard the specimen in the microscope with an incident angle of 45 degrees to the electron beam axis. This paper describes the outline of the facility and some of the recent results mainly concerned with heavy radiation damage of materials which are relevant to fast breeder and fusion reactor development. For example, we have investigated microstructural evolution of SUS 316 stainless steel as a function of dose, dose rate and temperature. The topics will also include observation of short-lived clusters of point defects during irradiation in nickel and direct comparison of self-ion damage in aluminum with electron damage caused by an electron beam within the 200 kV microscope. Some of these results have been discussed, compared with cascade simulation and defect kinetics calculations. The experimental results may be useful to establish correlation between neutron and ion damage through microstructural evolution modelling.

IASCC Susceptibility and Slow Tensile Properties of Highly-irradiated 316 Stainless Steels

Irradiation assisted stress corrosion cracking (IASCC) of cold-worked 316 stainless steels irradiated to doses up to 53 dpa was examined using slow strain rate tensile tests in 593 K simulated pressurized water reactor primary water while changing the content of dissolved hydrogen (DH) and dissolved oxygen (DO). A higher susceptibility was observed for higher doses and DH content, accompanied by increased corrosion product formation on the fracture surface and higher hydrogen accumulation near the fracture surface. At 53 dpa the susceptibility at both 0.02 and 8ppm DO was comparable to that at high DH content. The results indicated that IASCC was sensitive to DH content at doses less than 35 dpa but was less sensitive to both DH and DO content at 53 dpa. The subcrack formation and hydrogen accumulation in the hydrogenated condition suggested that processes associated with hydrogen would have an important role in IASCC in hydrogenated water. The same stainless steels were susceptible to intergranular type fracture during slow tensile tests in pure argon at 593 K. The intergranular type region consisted of a mixture of intergranular and dimple regions and the intergranular fraction was much smaller than that in IASCC in water environment.

Deformation Structure in Highly Irradiated Stainless Steels

Surface steps and deformation microstructure in cold-worked SUS316 stainless steels irradiated to 4 and 35 dpa (displacements per atom) were examined after being deformed by uniaxial tensile stress at 320°C at a slow or fast strain rate. Dislocation channeling was the predominant mode of deformation near the surface at the slow strain rate. Twinning was dominant at the fast strain rate whereas twinning and nanotwin formation occurred in the locally stressed area at the slow strain rate. Deformation heterogeneity measured using the spacing of coarse surface steps induced by dislocation channels increased with increasing dose from 4 to 35 dpa. Grain boundary separation occurred when dislocation pileups and high normal stress on the grain boundary plane coexisted, which likely was a precursor of intergranular cracking without any environmental factor.

Separation of Microstructural and Microchemical Effects in Irradiation Assisted Stress Corrosion Cracking using Post-irradiation Annealing

Post-irradiation annealing (PIA) was conducted in order to clarify the role of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) susceptibility in simulated pressurized water reactor (PWR) primary water. Microstructures, hardening, radiation-induced segregation and IASCC susceptibility were examined in cold-worked SUS 316 stainless steels irradiated to 25 dpa in a PWR after annealing at 673–823 K for 1h. IASCC susceptibility, microstructures and hardening recovered as the annealing temperature increased whereas the grain boundary segregation of Cr and Ni remained almost unchanged. The results suggested that the change in IASCC susceptibility due to annealing is not attributed to the change in grain boundary segregation but to the change in micro-structures and hardening. The fact that a smaller recovery of radiation hardening caused a larger IASCC susceptibility suggested that a threshold hardening level exists for the occurrence of IASCC.

Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking

Isolation of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) was attempted by means of low-dose high-temperature neutron irradiation in a material test reactor to get better understanding on IASCC. Microstructure, grain boundary segregation, hardness and SCC susceptibility were examined on stainless steels irradiated to 0.8 dpa at around 673 K. The irradiation caused well-developed grain boundary segregation without notable hardening or microstructural changes. No IASCC was found in 593 K hydrogenated water whereas the steels were highly susceptible to IASCC in 561 K oxygenated water. The results suggest that grain boundary segregation, probably Cr depletion, is sufficient to cause IASCC in oxygenated water and that other radiation-induced changes such as microstructure and hardening are required for IASCC in hydrogenated water.

In-situ microstructural observation of radiation damage in nickel produced by energetic heavy particles

Abstract Microstructural changes during 300 and 400 keV Ar+ irradiation in pure nickel between 300 and 773 K have been observed in-situ in an electron microscope. Some of the observations are recorded on a video tape. Various phenomena characteristic of cascade damage have been observed. Clustering of point defects is influenced strongly by the presence of point defects sinks: surfaces, pre-existing dislocations, loops and cavities. Wedge-shaped specimens are utilized to sort out the complex behavior of microstructural evolution. Of great interest is the fact that under certain conditions, metastable defect clusters with a very short lifetime are formed during irradiation at 773 K. The implication of these observations to fusion neutron damage modeling is discussed.

IASCC Initiation in Highly Irradiated Stainless Steels under Uniaxial Constant Load Conditions

Uniaxial constant load tests in 320°C simulated PWR primary water were conducted for cold-worked SUS316 stainless steels irradiated to 30–73 dpa. IASCC failure occurred within ∼100 h at applied stresses of 470–752MPa. The time-to-failure decreased with increasing neutron dose and applied stress. A compilation of literature data obtained using ring compression tests indicated that the results of ring compression tests for IASCC initiation were similar to those of uniaxial tensile tests. The border stress for IASCC failure decreased with increasing dose and saturated at ∼400 MPa at ∼30 dpa. This behavior was similar to that of radiation-induced material changes, such as hardness or grain boundary segregation.