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Reactor pressure vessel (RPV) irradiation embrittlement has raised increasing concern again since the middle of the 1990s. One aspect was the beginning discussion about possible long term operation beyond the plant lifetime. Another aspect was the end of the Cold War together with the opening of Eastern Europe. The RPV of WWER-440 plants located in Eastern Europe are known to be susceptible to irradiation embrittlement because of their geometry leading to a high neutron flux. Due to both aspects, many research projects have been started in the EU and worldwide to investigate this ageing phenomenon being of central importance to light-water reactor (LWR) plants.

The mechanism of irradiation embrittlement has been known since the early days of nuclear power plants, yet there were no physical models allowing quantitative predictions. Thus, surveillance programmes were established within the RPV to monitor its embrittlement for the design lifetime using irradiation specimens. The mechanism is well understood. However, details of the mechanism and improving the predictions still remain under investigation.

1.1.1 Brief Description of Mechanism

Irradiation embrittlement is the significant loss of fracture toughness and ductility of a material due to neutron and gamma irradiation. In nuclear power reactors this effect is generally dominated by neutrons. In principal it affects all kinds of metallic materials, yet with different dependence on neutron fluence (i.e. the neutron dose), energy spectrum of the neutrons, and temperature. In a hard sphere model of the metal lattice the displacement of atoms from their original lattice position due to collisions with neutrons causes a so-called displacement cascade resulting in vacancies and atoms at interstitial positions. Most defects recombine with interstitials, but a small number survives and migrates throughout the metal. Diffusion processes within the material lead to agglomerations of point defects like dislocation loops and microvoids. This matrix damage together with very fine precipitates of chemical composition differing from the matrix material as well as grain boundary segregation of accompanying elements (“impurities”) are considered to play major roles during embrittlement [Keim2012] [IAEA98] [IAEA05a] [IAEA05b] [IAEA07a] [IAEA07b] [IAEA09] [PMMD07].

The radiation damage depends on the number of atom displacements inducing by neutrons. In steels most displacements are created by fast neutrons, but thermal neutrons also contribute to the embrittlement process. The damage created by neutron irradiation may hence be expressed in displacements per atom (dpa) or as the number of neutrons exceeding a certain energy level (typically E > 1 MeV or E > 0.5 MeV if the RPV is concerned). The neutron spectrum plays a role for the radiation damage, but the influence is not quantitatively understood. Thus, the comparison between data obtained in different reactor environments, e.g., between research / test reactor and LWR environment must be done carefully [PRIS2004]. Relations between the different neutron damage scales depending on reactor environment have been established to compare data coming from different research facilities. In addition to irradiation embrittlement, thermal ageing can also contribute to embrittlement [JRC46534] [JRC63603] [Debarberis1998].

In ferritic materials like carbon and low-alloy steels, which typically exhibit a sharp ductile-to-brittle transition temperature (DBTT), neutron irradiation shifts this transition temperature to higher temperatures and decreases the level of toughness in the ductile regime (the “upper shelf”) [JRC46534] [Horvath2005]. The effect saturates at a certain neutron fluence level which depends on material and environment. Typical values are 1021 – 1022 n/cm2 (E > 1 MeV). Austenitic stainless steels do not exhibit such a pronounced DBTT and are much less sensitive to irradiation embrittlement. At high levels of neutron fluence, however, they also show a significant decrease of toughness and ductility [IAEA09] [PMMD07].

1.1.2 Information on NPP components

Irradiation embrittlement is a significant ageing mechanism in nuclear power reactors for the cylindrical shells of the RPV close to the core (the “RPV beltline”), RPV internals, and core components. Irradiation embrittlement mainly raises concern with regard to the integrity of the RPV because the integrity of the RPV as the centrepiece of a NPP is of utmost importance and must be guaranteed under all circumstances. Thus, extensive research has been triggered about the effect of neutron irradiation on typically used RPV steels. In particular welds and their heat affected zones in the beltline region are in the focus as they are considered to be more likely to contain cracks or defects than base metals and in many cases turned out to be more sensitive to irradiation embrittlement [JRC46534].

In the case of the LWR RPV, specimens made from the ferritic base and weld metals of the RPV beltline are usually irradiated during operation with a higher neutron flux than the rest of the RPV. This lead factor allows the prediction of the material behaviour before the same level of neutron fluence and hence of embrittlement in the RPV wall is reached. In general, there is a lack of reliable data necessary for long term operation in LWR environment, especially for austenitic steels. In particular, there is concern about the (ferritic) shell beltline of pressurized water reactor (PWR) RPV with its respective welds and nozzles depending on the choice of material, especially with regard to long term operation [IAEA05a] [IAEA05b] [IAEA07a] [IAEA07b] [IAEA09] [PMMD07]. Irradiation embrittlement is also of importance for CANDU pressure tubes [IAEA98].

1.1.3 Topics of Research Projects

For the last decade, interest in irradiation embrittlement has been motivated by the increasing demand for long term operation of a large number of NPPs. In this context, safety margins in design rules have been reviewed to reconsider effects of irradiation embrittlement during design life and long term operation in a state-of-the-art approach to RPV integrity.

There is much less focus on irradiation embrittlement of austenitic stainless steels used for RPV cladding, RPV internals, and core components. In a conventional structural analysis, the relatively thin RPV cladding is not considered as part of the pressure retaining boundary and therefore its properties are mainly important with respect to the corrosion protection of the RPV base metal. Besides, as the austenitic cladding is less prone to irradiation embrittlement as the ferritic material, changes in material properties are less significant. In contrast to cladding, RPV internals and core components within the core and close to it are exposed to irradiation levels high enough to cause significant reductions in toughness and ductility. Yet, they are subject to relatively low tensile stresses making these reductions less important [IAEA05a] [IAEA05b] [IAEA07a] [IAEA07b] [IAEA09] [PMMD07].

The research funded / supported by the EU in the Framework Programmes 4, 5 and 6 and the JRC Direct Actions is mostly focused on RPV materials of PWR reactors with emphasis on WWER reactors (Russian type PWR). Some investigations were performed on Magnox materials. The portability of the gained results to the advanced gas-cooled reactor (AGR) must be critically reviewed due to different operating conditions and geometrical boundaries. Few investigations were performed on boiling water reactor (BWR) materials and BWR issues. Very few studies are available about CANDU materials and specific CANDU issues.