Aki Toivonen

Current development on advanced technology fuel (ATF) claddings is aiming at improved high temperature integrity of new candidate materials designed on existing zirconium cladding materials. To assess their performance in commercial use, characterization of newly proposed materials is essential. The primary requirement for the cladding materials is the ability to tolerate loss of cooling for a significant period without failing. The tests in this work were performed on different types of coated Zr-alloys in a high temperature steam furnace in flowing steam conditions at 1100 °C for 500 s, 1200 °C for 1800s, and 1300 °C for 300 s. In addition, autoclave exposures were performed in simulated pressurized water reactor (PWR) water chemistry to confirm the material viability in normal light water reactor (LWR) operating conditions. After autoclave and steam furnace tests, exposed specimens were characterized using a Zeiss Crossbeam 540 field emission gun-scanning electron microscope (FEG-SEM) equipped with a semiquantitative energy dispersive X-ray spectrometer (EDS). Oxide and coating layers were imaged using secondary electrons and elemental composition of the layers were determined using selected area (SA) and point analyses. In addition, elemental maps were collected from specific areas of interest. Most of the tested specimens indicated detached coating layer. Varying amounts of cracking in the coating materials were present. Some of the cracks extended even into the base material. Based on this study, further developments of ATF cladding solutions are needed through international collaboration.

This paper describes the methods and results of analytical TEM examinations and autoclave testing of two highly-irradiated flux thimble tube materials harvested from a commercial pressurized water reactor. The materials are cold-worked 316L, and accumulated 65 dpa and 100 dpa of radiation dose. To set the baseline for a broader study, the materials were examined in the as-irradiated condition and tested as O-ring specimens at relatively high constant loads in simulated PWR water conditions. Tests were also conducted with elevated hydrogen. For a given load, more rapid cracking was associated with higher radiation dose, and with the elevated hydrogen.

The efficiency of ultra-supercritical (USC) steam power plants is limited by the materials properties, in particular, the steam oxidation resistance of the currently used steels at temperatures higher than 600 °C. Under these conditions, steam oxidation results in the development of thick oxide scales which spall and can accumulate in tube bends leading to blockage, overheating and premature creep rupture, as well as erosion of downstream components such as steam valves and turbine blades. Most published work related to oxidation testing is carried out at atmospheric pressure, with significantly less testing of austenitic steels in supercritical steam, and rarely including protective coatings. Indeed, the effect of high-pressure steam in the oxidation process is not quite understood at present. This paper covers a comparison of the behaviour of TP347HFG stainless steel at 700 °C under atmospheric pressure and 25 MPa, with and without slurry-applied diffusion aluminide coatings. The results show a very protective behaviour of the aluminide coatings, which develop a very thin Al-rich protective oxide, and no significant difference between the two environments. In contrast, the uncoated steel exhibited a different behaviour. Indeed, under atmospheric pressure after 3000 h, very thin scales, rich in Cr and not surpassing 5 to 10 µm in thickness, covered the samples along with some much thicker Fe-rich oxide nodules (up to 150 µm). However, under 25 MPa, a thick multilayer scale with a non-homogeneous thickness oscillating between 10 to 120 µm was present. A microstructural investigation was undertaken on the oxidised uncoated and coated substrates. The results suggest that pressure increases the oxidation rate of the chromia former steels but that the oxidation mechanism remains the same. A mechanism is proposed, including early detachment of the outer growing scales under supercritical pressure.

The presented work consists of a test setup study of a new pneumatic material testing device based on double bellows (DBs) loading device and with miniature autoclaves enabling applications at temperature and pressure up to 650 °C and 35 MPa, respectively. It has been demonstrated that it is technically feasible to carry out well defined and controlled material testing in the supercritical water (SCW) environment using this testing system. By using this type of system, it makes possible to investigate the intrinsic role of the applied stress on the deformation behavior of material in light water reactor (LWR) conditions and also in other harsh environments like SCW conditions. In addition, the compactness and versatility of the setup makes this system particularly attractive for deployment in a hot-cell for testing of irradiated materials.

Environmentally assisted cracking (EAC) susceptibilities of Alloys 600, 690, and their weld metals with and without weld defects were benchmarked in pure and doped hydrogenated steam at 400°C and pressurized water reactor (PWR) water at 360°C on four-point bend (0.35% to 1.1% strain) and U-bend specimens (5% strain). Contrary to the expectations, no EAC initiation from existing weld defects was observed. As expected, Alloy 600 and its weld metals were much more prone to EAC initiation than Alloy 690 and its weld metals. Oxide structures remain the same in doped and pure steam and PWR water, although the environment has an impact on the EAC initiation times and oxidation rates. Because of the similar oxide structures, the EAC mechanism is assumed to be the same in all studied environments.

The kinetic parameters of oxidation of two oxide dispersion strengthened (ODS) alloys, PM2000 and MA956, in supercritical water (SCW) are evaluated using an updated model that assumes that the growth of the outer layer is governed by the transport of interstitial cations through the inner layer. The model is able to reproduce quantitatively the depth profiles of individual constituent elements in the inner and outer layers, as well as in the diffusion/transition layer of the alloy between the inner layer and the bulk substrate. The rate constants and diffusion coefficients decrease with time, indicating oxide layer restructuring.