Wednesday, April 22, 2015

The IASCC in Steels and the example of How Materials Science can support the Nuclear Development

A nuclear reactor is a very harmful environment for materials: neutrons with no electric charge can penetrate into alloys and crystalline networks promoting displacements of its atoms from equilibrium positions. This mechanism is responsible to change material's properties in which could result in a nuclear accident by failure of internal components. The task of choosing materials to operate and compose the structure of a nuclear reactor has paramount importance in nuclear activity, notwithstanding this is a major challenge.

Advanced Stainless Steels has been studied for nuclear reactors internal components because its good properties: relatively high strength, ductility, fracture and corrosion resistance. But in the middle of eighties, when IASCC (Irradiation-Assisted Stress Corrosion Cracking) was discovered and deeply studied, the steels were phased out in light water nuclear applications for safety reasons.

What is the IASCC? The scientific knowledge regarding Stress Corrosion Cracking (SCC) for materials operating in high- temperature and -pressure conditions depends on three main issues: high-stress, extreme harmful place and a susceptible material. As aforementioned, neutron can produce damage in materials by displacing atoms in equilibrium positions and the material will loose its ductility. The embrittlement caused by neutron irradiation will contribute to the evolution of crack tips in the material. So then IASCC is a mechanism of crack growth/formation in a corrosive environment assisted by neutron irradiation. Steels are aggressively affected by the IASCC [1].

Unfortunately, there are no data regarding crack evolution and growth mechanisms in Steels by the IASCC in nuclear reactors. Nowadays, failures of internal components are reported in nuclear reactors for fluences around 5E22 neutrons per centimetre-square and there is no strong correlation between IASCC and failure of PWR or BWR internal components. There is a lack in science and nuclear engineering here, therefore metallurgy and materials sciences should be more addressed to face  and solve problems in nuclear field: this will enhance the safety of nuclear materials in harmful environments and will improve the overall operation of a nuclear reactor.


[1] O.K. Chopra, A.S. Rao, A review of irradiation effects on LWR core internal materials – IASCC susceptibility and crack growth rates of austenitic stainless steels, Journal of Nuclear Materials, Volume 409, Issue 3, 28 February 2011, Pages 235-256.

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