Nuclear Power – Control, Reliability and Human Factors, Pavel Tsvetkov (Ed.), ISBN: 978-953-307-599-0, InTech
Diego Ferreño, Iñaki Gorrochategui and Federico Gutiérrez-Solana
The pressure vessel constitutes the most important structural component in a nuclear reactor from the point of view of its safety. The core of the reactor, that is, the nuclear fuel, is accommodated inside the vessel. This material is composed of fissile nuclides that undergo chain nuclear reactions of an exothermic nature, thus generating usable energy. Uranium 235 (U-235) is the only isotope in Nature which is fissile with thermal neutrons; hence, it is used as nuclear fuel in Light Water Reactors (LWRs). Two technologies of LWRs can be distinguished, Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). Currently, more than 400 nuclear reactors operate in the world of which, approximately, 57% are PWR and 22% BWR. The original design lifetime for LWRs is 40 calendar years; nevertheless, the current target for most plants in many countries in Europe, Japan and the USA is to extend their operative lifetime up to 60 years.
The nuclear vessel is a virtually irreplaceable element which is subjected to operating conditions that lead to a progressive degradation in time of its constituent steel. The chain fission reactions of U-235 entail the emission of high energy neutrons that inevitably impact the internal surface of the vessel. These collisions give rise to a complex series of events in the nano and microstructural scale that, in the end, modify the mechanical properties of the steel leading to its embrittlement, that is, the decrease in its fracture toughness. This phenomenon is most intense in the so called beltline region (which is the general area of the reactor vessel near the core midplane where radiation dose rates are relatively high). The total number of neutrons per unit area that impact the internal surface of the vessel represents the neutron fluence; in practice, only that fraction of the energy spectrum corresponding to a neutron kinetic energy higher than 1 MeV is considered to be capable of triggering damage mechanisms in the vessel steel; these neutrons are referred to as fast neutrons. Typical design end of life (EOL) neutron fluences (E>1 MeV) for BWRs are in the order of 1018 n/cm2, whereas for PWRs this number is about 1019 n/cm2. In Section 2 of this chapter, the embrittlement of nuclear vessel steels is described from a purely phenomenological perspective as well as from the point of view of the legislation currently in force. The phenomenon of the ductile to brittle transition and the influence of embrittlement on it are particularly stressed. In Section 3, a brief description of the main characteristics of the nuclear power plants surveillance programmes is presented; the requirements that they envisage as well as the information that they allow to be obtained are pointed out. The physical mechanisms that take place in the nano and micro levels leading to the material embrittlement are detailed in Section 4 where a brief exposition concerning the most relevant predictive models for embrittlement is also presented. Finally, in Section 5, the promising method of the Master Curve is described; this represents an improved methodology for the description of the fracture toughness of vessel steels in the ductile to brittle transition region, susceptible to be incorporated in the current structure of the surveillance programmes.