Application of the FRI crack growth model for neutron-irradiated stainless steels in high-temperature water of a boiling water reactor environment

Koshiishi, Masato / Hashimoto, Tsuneyuki / Obata, Ryoji | British Library Online Contents | 2017

Intergranular Stress Corrosion Crack Growth of Sensitized Type 304 Stainless Steel in a Simulated Boiling-Water Reactor Environment

Kikuchi, E. / Itow, M. / Kuniya, J. et al. | British Library Online Contents | 1997

Factors Affecting the Crack Growth Rates of Reactor Pressure Vessel Steels Under Simulated Boiling Water Reactor Conditions

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Corrosion fatigue crack growth behaviour of low-alloy reactor pressure vessel steels under boiling water reactor conditions

Seifert, H. P. / Ritter, S. | British Library Online Contents | 2008

Fatigue Crack Growth Behavior of Sensitized Type 304 Stainless Steel Under Boiling Water Reactor Conditions

Chen, Y. Y. / Shih, H. C. / Wang, L. H. et al. | British Library Online Contents | 2007