Publications Repository - Helmholtz-Zentrum Dresden-Rossendorf

1 Publication
Previous results of the investigation on the decommissioned reactor pressure vessels of the Greifswald NPP
Viehrig, H.-W.; Houska, M.;
The investigation of reactor pressure vessel (RPV) material from the decommissioned Greifswald NPP representing the first generation of Russian type WWER-440/V-230 reactors offers the opportunity to evaluate the real toughness response. The Greifswald RPVs represent different material conditions viz. irradiated, irradiated and annealed and irradiated, annealed and re-irradiated.
The paper presents test results measured on the trepan taken from the beltline welding seam located in the reactor core region of the Unit 4 RPV. This unit was shut down after 11 years of operation and represents the irradiated condition. The working program comprises chemical analysis, microstructure investigations (by means of metallography and SEM), mechanical testing (hardness and tensile), and fracture mechanics testing. The key part of the testing is focused on the determination of the reference temperature T0 following the ASTM test standard E1921-10. The KJc values measured on TS oriented pre-cracked and side-grooved Charpy size SE(B) specimens from defined thickness locations of the welding seam approximately follow the course of the Master Curve but with a large scatter. Reference temperatures T0 through the thickness of the RPV beltline welding seam runs almost opposite to the trend predicted by the Russian code for the decrease of the neutron fluence from 5.1•E19 n/cm2 to 1.1•E19 n/cm2 (E>0.5MeV). T0 varies between 6°C from near the welding root at ¼ wall thickness to 117°C at ¾ wall thickness. The scatter of T0 beyond the welding root is about 40 K and depends strongly on the structure at the crack tip. The results are also evaluated according to integrity assessment procedures based on the Master Curve concept.
Keywords: reactor pressure vessel, welding seam, specimen orientation, fracture toughness, Master Curve approach, integrity assessment
  • Lecture (Conference)
    IAEA Specialists’ Meeting on Irradiation Embrittlement and Life Management of Reactor Pressure Vessels, 18.-22.10.2010, Znojmo, Czech Republic

Publ.-Id: 14825 - Permalink