The aim of this work was to rationalize the corrosion resistance of candidate austenitic iron-chromium-nickel (Fe-Cr-Ni) alloys in supercritical water (SCW) for use as fuel claddings within the in-core structure of the Canadian supercritical water-cooled reactor (SCWR) concept. High chromium (Cr)-containing alloys (Alloy 800HT with 20.6 wt.% Cr and 30.7 wt.% Ni and Alloy 33 with 33.4 wt.% Cr and 31.9 wt.% Ni) in the mill annealed condition were chosen for this purpose. Coupons were exposed on a short-term basis (500 h) in a static autoclave containing 25 MPa SCW at 550 °C and 625 °C. Gravimetric measurements and electron microscopy techniques were then used to study the oxidation/corrosion resistance of two alloys. Alloy 33 was found to exhibit the higher corrosion resistance at both temperatures. The improved corrosion resistance of Alloy 33 was attributed to two factors: (i) the formation of a continuous Cr-rich corundum-type M2O3 (M= Cr and Fe) oxide layer that prevented the diffusion of Fe and the formation of a less-protective Fe/Mn-Cr spinel ([Fe,Mn]Cr2O4) outer layer, (ii) a sufficient residual bulk Cr in the Cr-depleted layer adjacent to the alloy/scale interface that prevented any localized internal oxidation from occurring. A mass balance conducted on the corroded Alloy 33 material suggested that volatilization of the corundum-type oxide layer did not occur, at least not within the short-term exposure in the essentially deaerated SCW. A key issue requiring further study was the observation of intermetallic precipitates that formed below the Cr-depleted zone adjacent to the alloy/scale interface in both alloys when exposed for 500 h at 625 °C and their possible influence on the in-service mechanical integrity. / Thesis / Master of Applied Science (MASc) / The supercritical water-cooled reactor (SCWR) is one of the six reactor design concepts developed by the Generation-IV International Forum (GIF). Canada is planning to build the SCWR within the next decades. However, selection of proper materials that perform well within such high pressure high temperature circumstances inside the reactor core with minimum degradation is a very imperative challenge. The current work has addressed this issue by studying the corrosion behaviour of Fe-Cr-Ni alloys in similar environment using electron microscopy techniques.

<p>The supercritical water cooled reactor (SCWR) is a Generation IV reactor concept that features light water coolant in a supercritical state. Canada is developing a pressure tube variant of the supercritical water reactor as an evolution of the CANDU reactor. The main advantages of the pressure tube SCWR are an improved thermal efficiency over current reactors, enhanced safety through passive safety features, and plant simplifications. The objective of this thesis was to investigate current fuel and core designs for the Canadian SCWR concept.</p> <p>Simulations of 2-D lattice cells for fuel assemblies containing 43 and 54 fuel elements were performed using the neutron transport code WIMS-AECL. Safety parameters and fuel burnup performance were investigated here. Three dimensional full core simulations were performed using the diffusion code RFSP. These studies examined batch fueling, cycle length, radial and axial power profiles, linear element ratings, and reduction of axial power peaking through graded enrichment along the fuel channel. Finally, a study of reactivity transients was performed using the FUELPIN heat transfer/point kinetics code.</p> <p>The main results of the studies show that the coolant density change that occurs as water passes through the pseudocritical point strongly affects fuel performance. It is concluded that the 54 element assembly design is acceptable in terms of coolant void reactivity performance with lattice pitch smaller than 26 cm. To meet the burnup target, a fuel enrichment of about 5% is required. From the RFSP studies, this level of fuel enrichment will provide an operating period of 370 days between refueling. Relatively high axial power peaking is observed at the beginning of cycle conditions. A main finding is that the proposed reactor power level of 2540 MWth produces unacceptably high linear element ratings. This is confirmed using the FUELPIN code. A reduction in linear element rating is suggested for consideration.</p> / Master of Applied Science (MASc)