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Warm Pre-Stress Tests of the Irradiated RPV Base Material of Beznau Unit 1

[+] Author Affiliations
Florian Obermeier, Stefan Heußner, Herbert Schendzielorz, Marco Kaiser, Elisabeth Keim

AREVA GmbH, Erlangen, Germany

Heinz Hägeli

Axpo AG NPP Beznau, Döttingen, Switzerland

Paper No. PVP2016-63511, pp. V06AT06A066; 10 pages
  • ASME 2016 Pressure Vessels and Piping Conference
  • Volume 6A: Materials and Fabrication
  • Vancouver, British Columbia, Canada, July 17–21, 2016
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 978-0-7918-5042-8
  • Copyright © 2016 by ASME


According to the pertinent regulations, the integrity of a reactor pressure vessel (RPV) of a nuclear power plant is to be assessed by fracture mechanics for postulated flaws under most severe loading conditions. In such an analysis usually loss of coolant accidents are assumed to cause highest possible loading of the structural material of the RPV. This is due to the fact that such a pressurized thermal shock (PTS) event during which cold emergency coolant is injected into the primary system generates additional thermal stresses in the RPV wall. Based on the applicable regulation, the initiation of postulated flaws is to be excluded by the comparison of the calculated crack tip loading and the fracture toughness of the particular material.

This kind of assessment was motivation of various research projects in the last decades addressing both evaluation approaches and experimental testing. A crucial result in this context is the existence of the so-called warm pre-stress effect (WPS) on the resulting fracture toughness. Generally, it is known as the increase of the apparent fracture toughness of a flaw in a specimen or structure after loading at high temperatures, generally in the upper shelf region, followed by a reloading at a lower temperature. This represents the typical loading scenario postulated for the assessment of a RPV during a PTS event.

Experiments were performed to quantify this effect in the case of the irradiated reactor pressure vessel base material of the nuclear power plant Beznau unit 1. This paper presents the results of the Master Curve tests to determine the reference temperature (according to ASTM E 1921) and the design and testing of the warm pre-stress experiments using irradiated 10×10 mm reconstituted single edge notch bend (SE(B)) specimens.

The design of these warm pre-stress tests was based on the loading transients for postulated surface and sub-surface flaws investigated within the scope of the assessment of the Beznau unit 1 reactor pressure vessel against brittle failure. Finite element simulations were performed to transfer the loading conditions at the crack tip of the RPV determined during the brittle fracture safety assessment onto the SE(B) specimen.

The simulation results were used to control the loading conditions as a function of time and temperature during the experimental tests. The fracture toughness values of the warm pre-stress specimens were finally compared with the original fracture toughness values determined in the absence of a warm pre-stress effect to demonstrate the increase of the safety margin when the warm pre-stress effect is taken into account.

Copyright © 2016 by ASME
Topics: Stress



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