Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Abstract

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

Oxidation kinetics of silicon carbide‐containing refractory diborides, I: Revised model

“…Indeed, bubble formation has been reported for oxidation of SiC in water vapor under conditions and partial pressures where no bubbles have been observed during oxidation by oxygen. [40][41][42][43] Note that the permeability of CO(g) in the scale growing during oxidation in or in the presence of water vapor may be different from that in dry oxygen. Hydroxyl content affects many properties of glass, including viscosity and gas permeation.…”

Radiation Effect in Ti-Cr Multilayer-Coated Silicon Carbide under Silicon Ion Irradiation up to 3 dpa

“…Reports about swelling of Ti and Cr at LWR normal operation temperatures are limited, while swelling of SiC has been researched a lot [22,55,68,69]. The swelling of SiC for the temperature regime from approximately 473 K to 1073 K is referred to as the point-defect swelling regime [55].…”

“…Since SiC has a lower hydrogen generation rate and lower heat of reaction than Zr, the amount of hydrogen generated from fuel claddings by reaction with steam during severe accidents is expected to significantly decrease when SiC fuel cladding is used instead of conventional Zircaloy fuel cladding [3,4]. Therefore, the expectation for increased accident tolerance by applying SiC fuel cladding has led to the development of manufacturing and integration technologies [5][6][7][8][9][10][11][12][13][14][15][16], research on material behavior under normal operation and accident conditions [5][6][7][17][18][19][20][21][22][23][24][25][26][27][28][29][30][31], and evaluation of safety and economic benefits [32].…”

“…Although the steam oxidation rate of the specimens is affected by pressure and flow rate of steam [19], the corrosion can be limited to the surface for the temperature range of 1973-2073 K in which formation of SiO 2 bubbles occurred on the surface by softening and melting of SiO 2 [20,22]. CVD-SiC-overcoated SiC/SiC composite tubes maintained a coolable shape after quench tests at 2273 K [21,22].…”

Effect of bias voltage on the microstructure and high‐temperature steam oxidation behavior of Cr coatings prepared by multi‐arc plating on Zircaloy‐4 alloy

“…So far, numerous types of coatings have been proposed by researchers for the surface modification of zirconium alloy nuclear fuel cladding tubes, mainly including metal coatings, [8][9][10] ceramic coatings, [11][12][13] and multicomponent composite coatings. [14][15][16] Among these coatings, the metallic Cr coating has good corrosion and oxidation resistance, excellent mechanical strength and wear resistance, favorable adhesion with zirconium alloy substrates, and a relatively low thermal neutron absorption cross-section, [17][18][19][20] which is considered the most promising candidate material for zirconium alloy cladding coating.…”