The surveillance of the reactor pressure vessel (RPV) material is not directly a countermeasure against irradiation embrittlement but may allow the use of more realistic values of material properties for the integrity assessment of the RPV than predictive formulas. The surveillance of light-water reactor (LWR) RPVs includes destructive testing of specimens made from the ferritic base and weld metals used for the RPV beltline which are irradiated during operation with a higher neutron flux than the RPV itself. This lead effect allows the prediction of the material behaviour before the same level of neutron fluence and hence of embrittlement in the RPV wall is reached.
During the VERLIFE project, procedures and guidelines were developed how to handle the integrity assessment and the plant lifetime calculation of WWER plants against the background of irradiation embrittlement of the RPV based on surveillance specimens, continuous neutron fluence measurements, and the end-of-life design fluence as a basis for evaluation. In addition, the project outcomes are seen as guidance for the periodic safety analysis / reports [Brumovsky2014].
When regarding long term operation like 60 or even 80 years of operation, new challenges of surveillance occur. Different strategies have been discussed within the LONGLIFE project to allow for irradiation embrittlement monitoring of this long term operation periods. Firstly, surveillance capsules could be placed closer to the core or modified to enhance the lead factor. Secondly, removed capsules could be reinserted for additional irradiation (potentially with using reconstitution techniques). Thirdly, new capsules could be manufactured using archive material specimens if available. Fourthly, the use of "surrogate" material could be an alternative in the case of missing original material specimens. When the operation time is 80 years or even beyond regions of the RPV outside the beltline may be affected by irradiation embrittlement. There may be concern if deleterious elements such as copper and phosphorous are present in higher concentrations in these materials. Usually, specimens of these materials are not available. Thus, new strategies such as the use of generic material properties or a restriction in pressure-temperature limits are seen to be feasible. As a result, guidelines for a Coordinated Reactor Vessel Surveillance Programme were developed to monitor irradiation embrittlement during life extension to 60 or 80 years [Ballesteros2014].
Annealing is seen as the most effective measure to reduce the effect of irradiation embrittlement or even fully recover the initial properties of the material. Extensive studies have been performed to find best values for temperature and annealing time in both Western and Eastern RPV steels. Typically, an annealing temperature around 150–200 °C above the operating temperature and an annealing duration of about one week is sufficient to recover mechanical properties of the RPV wall. While annealing was rather seldom performed in Western PWR in operation due to geometrical boundary conditions and concerns about possible negative side effects of annealing on the microstructure of RPV steels, the beltline welds of the RPV of many WWER-440 plants have been thermally annealed during extended refuelling outages [IAEA09]. For the annealing process, the core and all internals are removed. Afterwards, the RPV is heated up using external electric heating devices and slowly cooled down after annealing. In this context, the efficiency of the annealing depends on the irradiation history, the annealing conditions, and the content of alloying and accompanying elements. The risk of thermal embrittlement during the annealing can be ruled out as far as the duration of the heat treatment does not exceed 1000 h [JRC46534]. Re-embrittlement after annealing is also captured by the semi-mechanistic model of Debarberis et al. as far as the database allows for [JRC30537].
Ulbricht et al. systematically studied the effect of annealing on two differently irradiated RPV steels. Each material was irradiated up to several doses at 255 °C in a WWER-2 prototype reactor. After irradiation, individual specimens of each material were annealed at temperatures between 350 and 475 °C for 10 h and investigated using the Small Angle Neutron Scattering technique. For both materials it was shown that with increasing annealing temperature the irradiation hardening effect can be – at least partly – healed-out [Ulbricht2006].
When investigating the influence of neutron irradiation on WWER-440 RPV cladding materials (ASS), Gillemot et al. revealed that irradiated-annealed samples, when re-irradiated, behave comparable to unirradiated specimens concerning the mechanical properties [JRC34527].
Semi-mechanistic analytical model for radiation embrittlement and re-embrittlement data analysis, 2005.
Radiation stability of WWER RPV cladding materials, 2007.
Annealing and re-embrittlement of reactor pressure vessel materials, State of the art report ATHENA WP-4, AMES Report N. 19, 2008.