Investigation of Beltline Welding Seam of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

Abstract

The investigation of reactor pressure vessel (RPV) materials from decommissioned nuclear power plants (NPP) offers the unique opportunity to scrutinize the irradiation behavior under real conditions. The paper describes the investigation of trepans taken from the decommissioned WWER-440 RPVs of the Greifswald NPP. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step, the trepan taken from the RPV Greifswald Unit 1 containing the multilayer welding seam located in the beltline region was investigated. This welding seam represents the irradiated, recovery annealed, and reirradiated condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. After an initial increase of T0 from 10°C at the inner surface to 49°C at 22 mm distance from it, T0 decreases to −32°C at a distance of 70 mm, finally increasing again to 61°C near the outer RPV wall. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on subsize Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature, TT41J, estimated with results of subsize specimens after the recovery annealing, was confirmed by the testing of standard Charpy V-notch specimens.

Small-angle neutron scattering study of neutron-irradiated and post-irradiation annealed VVER-1000 reactor pressure vessel weld material

“…This kind of benefit is based on a partial or complete recovery of both the as-irradiated nanostructure and the mechanical properties of the RPV material. In this context, studies of the annealing behavior formed the basis to identify optimum post-irradiation annealing regimes in terms of temperature and time (Popp et al, 1989) (e.g., 475 °C/152 h for VVER-440 reactors) or to demonstrate the success of realized high-temperature dry annealings (Pelli and Törrönen, 1998;Viehrig et al, 2009). Alternatively, less demanding and more cost-efficient procedures such as low-temperature (e.g., 343 °C) wet annealing within the RPV design limits (Pelli and Törrönen, 1998;Krasikov, 2012) might give rise to partial recovery possibly allowing for some limited but worthwhile lifetime extension.…”