Materials Challenges in Nuclear Energy

Review of supercritical CO2 power cycle technology and current status of research and development

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

Recent progress in understanding dislocation interactions with grain boundaries and interfaces in metallic systems via static and in situ dynamic experimental approaches is reviewed.

Dislocation interactions with grain boundaries

The focus of this review is the state of knowledge of the oxidation behaviour in steam of alloys with potential for use as pressure parts in steam boilers. The rate of oxide growth on steam-touched surfaces and the characteristics of that oxide are of increasing interest as the quest for improvements in cycle efficiency leads to progressively higher operating temperatures and pressures. The consequences of increased rate of growth of these oxides are of concern because of implications for tube overheating and oxide exfoliation. Mitigation of such problems requires a mechanistic understanding of the influences of alloy composition and microstructure, and especially of the evolution with time of specific scale structures. Similarly, the relative effects of factors such as time, temperature and operating parameters must be understood. The oxidation behaviour of the class of ferritic steels that forms the bulk of the heat transfer surface in steam boilers is of particular importance since allo...

A review of the oxidation behaviour of structural alloys in steam

Corrosion and stress corrosion cracking in supercritical water

IGSCC of grain boundary engineered 316L and 690 in supercritical water

The Supercritical water-cooled reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification compared to current light water reactors (LWRs). Reactor operating conditions call for a core coolant temperature between 280 and 620 °C at a pressure of 25 MPa and maximum expected neutron damage levels to any replaceable or permanent core component of 15 dpa (thermal reactor design) and 100 dpa (fast reactor design). Throughout the core, corrosion, stress corrosion cracking, and the effect of irradiation on these degradation modes are critical issues. This chapter reviews the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-offs associated with certain alloy systems.

E.A. West

Papers

Corrosion and stress corrosion cracking in supercritical water

IGSCC of grain boundary engineered 316L and 690 in supercritical water

Material Performance in Supercritical Water

A model for the normal stress dependence of intergranular cracking of irradiated 316L stainless steel in supercritical water

Effect of irradiation on stress corrosion cracking in supercritical water