IN view of the interest attaching to the vaporisation and diffusion of solids, the following observations may be worthy of record.

The Diffusion of Solids

Passivity of 316L stainless steel in borate buffer solution studied by Mott-Schottky analysis, atomic absorption spectrometry and X-ray photoelectron spectroscopy

Uniform corrosion of FeCrAl alloys in LWR coolant environments

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.

https://www.tandfonline.com/doi/pdf/10.1080/00223131.2013.772448

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

In this study, we report that liquid nitrogen treatment is a promising zero-valent iron activation method that does not remove the iron oxide shell; this can improve the apparent Cr(VI) removal rate constant of zero-valent iron by about 4-120 times, depending on the particle sizes and the suppliers of zero-valent iron. It was found that liquid nitrogen, with its low temperature of 77 K, could crack the iron oxide shell of zero-valent iron to produce abundant fractures because of the different thermal expansion coefficients of iron oxide and iron. These fractures provided suitable mass transfer channels for the inward transfer of water/oxygen molecules to the iron core and the subsequent in situ generation of Fe(II) for the reduction of Cr(VI) to Cr(III). More importantly, systematic characterizations confirmed the generation of an Fe(III)/Cr(III)/Cr(VI) composite on the surface of zero-valent iron during the removal, suggesting its environmental benignancy. This study provides a novel physical zero-valent iron activation method, sheds light on the importance of the iron oxide shell of zero-valent iron on Cr(VI) removal, and clarifies the intrinsic Cr(VI) removal mechanism of zero-valent iron.

Liquid Nitrogen Activation of Zero-Valent Iron and Its Enhanced Cr(VI) Removal Performance.

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 barri...

https://www.tandfonline.com/doi/pdf/10.1080/18811248.2008.9711883

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

The growth rate of stress corrosion cracking (SCC) was measured for cold-worked, thermally treated Alloy 690 (UNS N06690, CW TT 690) and cold-worked, solution-treated Alloy 690 (CW ST 690) in hydrogenated pressurized water reactor (PWR) primary water under static load condition. Three important patterns were observed. First, intergranular stress corrosion cracking (IGSCC) was observed in CW TT 690 in PWR primary water in the range between 320°C and 360°C; this rate of SCC growth was slower than in CW mill-annealed Alloy 600 (UNS N06600, CW MA 600). No significant IGSCC was observed in CW ST 690 after 5,109 h in hydrogenated PWR primary water at 360°C. This is opposite of the behavior reported in the literature for high-temperature caustic solutions. Second, to assess the role of creep, rates of creep crack growth were measured in air, argon, and hydrogen gas environments using 20% CW TT 690 and 20% CW MA 600 in the range between 360°C and 460°C; intergranular creep cracking (IG creep cracking) wa...

Dependence of Stress Corrosion Cracking of Alloy 690 on Temperature, Cold Work, and Carbide Precipitation—Role of Diffusion of Vacancies at Crack Tips

SCC growth behaviors of austenitic stainless steels in simulated PWR primary water

The growth rate of stress corrosion cracking (SCC) was measured for nonsensitized and sensitized, cold-worked Type 316 (UNS S31600, CW316) and Type 304 (UNS S30400, CW304) in hydrogenated ...

Dependence of Stress Corrosion Cracking for Cold Worked Stainless Steel on Temperature and Potential and Role of Diffusion of Vacancies at Crack Tips

The purpose of this research was to compare the stress corrosion cracking (SCC) resistance of materials used at the present time for steam generator (SG) tubing in pressurized water reactor (PWR) primary water. Our results in PWR primary water for 20% cold-worked (CW) Alloy 800 (UNS N08800) are compared with already published data for SCC growth from 20%CW Alloy thermally treated (TT)690 (UNS N06690), 20%CW Alloy mill-annealed (MA)600 (UNS N06600), and 20%CW austenitic stainless steels. The second purpose was to examine the dependence of SCC growth on nickel and chromium in PWR primary water; the objective was to obtain the basic knowledge to understand SCC behavior of SG tubing materials. The third objective was to understand whether accelerated testing at higher temperatures is appropriate for predicting SCC initiation and growth at lower temperatures. For these objectives, SCC growth was measured in PWR primary water at 290, 320, 330, 340, and 360°C under static load conditions. Tests were performed us...

Tomoki Miyamoto

Papers

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

Dependence of Stress Corrosion Cracking of Alloy 690 on Temperature, Cold Work, and Carbide Precipitation—Role of Diffusion of Vacancies at Crack Tips

SCC growth behaviors of austenitic stainless steels in simulated PWR primary water

Dependence of Stress Corrosion Cracking for Cold Worked Stainless Steel on Temperature and Potential and Role of Diffusion of Vacancies at Crack Tips

Intergranular Stress Corrosion Cracking Growth Behavior of Ni-Cr-Fe Alloys in Pressurized Water Reactor Primary Water