RADIATION RESPONSE AND MECHANICAL PROPERTY CHANGES OF CHROMIUM AS ACCIDENT-TOLERANT FUEL CLADDING COATING

Abstract

The main goal is of this study is to surrogate the performance and identify possible challenges of using pure chromium as an accident tolerant fuel coating for existing Zircalloy nuclear fuel cladding. After the 2011 tsunami and subsequent loss of coolant accident in the Fukushima Daichi nuclear power plant a large emphasis has been placed throughout the nuclear community to find a mitigation and prevention of similar accident that are caused by the shortcomings of current zirconium based nuclear fuel cladding. Chromium has been chosen as the subject of this study as it is a prime contender for the usage as an accident tolerant fuel cladding coating due to its ability to withstand high temperature corrosion in aqueous environments, wear resistance and ease of application to already existing fuel cladding. Due to the lack of usage of chromium in its pure form in the nuclear industry little information is available on the performance of it under severe radiation conditions encountered in nuclear reactors. This work is divided into 3 parts to investigate necessary performance data for chromium. In the first part a series of 5 heavy ion (Fe) irradiations at 5 MeV between the temperatures of 450 °C and 650 °C are conducted to a damage levels of 50 dpa and damage rate of 1.75*10^(-3) dpa/s aimed at finding the peak swelling temperature of chromium and subsequently increasing damage levels to 100 and 150 dpa at the peak swelling temperature to showcase the swelling behavior of chromium. During this first phase void ordering is observed at high damage levels and steady state non saturated swelling of 0.04 -0.05 %/dpa are found after all data is combined. In the second part of this study the temperature range is increased to 6 temperatures between 350 °C and 650 °C and irradiations at damage rates of 3.5*10^(-3) dpa/s , 3.5*10^(-4) dpa/s and 3.5*10^(-5) dpa/s were conducted at each temperature to a fuel relevant peak damage level of 15 dpa. Results show a decrease in peak swelling temperature with increase in damage rate. Applying a relationship of the style 1/T=a*〖log〗_10 (K)+c as proposed by Brailsford and Bullough in 1972 enables the extrapolation of the chromium peak swelling temperature to reactor typical damage rates of 1*10^(-6) dpa/s to 1*10^(-7) dpa/s yielding peak swelling temperatures of 424 °C and 398 °C respectively. In the last part of the study a 2 MeV proton irradiation of pure chromium is performed, and the increasing damage profile is used to manufacture 4 micro pillars from the same grain with damage levels of 0, 0.5, 0.7 and 1 dpa respectively. Subsequent compression of the pillars sowed an increase of up to 57 % in resolved critical stress after 1dpa damage.