The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

• Published in: Journal of nuclear materials Vol. 350; no. 3; pp. 208 - 220
• Main Authors: Tsang, D.K.L., Marsden, B.J.
• Format: Journal Article
• Language: English
• Published: Amsterdam Elsevier B.V 01.05.2006; Elsevier
• Subjects: Applied sciences; Energy; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Installations for energy generation and conversion: thermal and electrical energy; United Kingdom; Europe; British Isles; M0200; C0100; M0300; Graphite moderated reactor; Stress analysis; Nuclear power plant; Graphite; Structure integrity; Neutron irradiation; Fast neutron; Oxidation; Gas cooled reactor; Stress strain; Nuclear reactor
• ISSN: 0022-3115; 1873-4820
• DOI: 10.1016/j.jnucmat.2006.01.015

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.