PSI - Issue 28

2

Versteylen/ Structural Integrity Procedia 00 (2020) 000–000

Casper Versteylen et al. / Procedia Structural Integrity 28 (2020) 1918–1929 © 2020 The Authors. Published by ELSEVIER B.V. This is an open access article under the CC BY-NC-ND license (https://creativecommons.org/licenses/by-nc-nd/4.0) Peer-review under responsibility of the European Structural Integrity Society (ESIS) ExCo Keywords: Cleavage; Reactor pressure vessel; CFD-FEM

1919

1. Introduction

A reactor pressure vessel needs to retain the integrity of the pressure boundary even in the most extreme loading conditions. In the event of a loss of coolant accident (LOCA), cold water can be injected in the still pressurized reactor pressure vessel (RPV). Pressurized thermal shock (PTS) can occur on the wall of a RPV due to the injection of cold cooling water in the still hot and pressurized RPV. The emergency core cooling system (ECCS) causes a temperature gradient in the RPV wall, which in turn causes thermal stresses. According to ASME codes, a postulated crack must be assumed in the RPV with a depth of 1/4 of the wall thickness and a length of 3/2 times the wall thickness. The thermal and hoop stresses are most severe for cracks along the axial direction of the RPV and the dominant fracture mechanism: mode I. The thermal stresses can be approximated for a specific thermal transient, but real thermal transients feature turbulent flows which leads to complex hot and cold spots and thermal gradients. The thermal stresses can therefore fluctuate in a similar fashion. A rigorous multidisciplinary approach is chosen in order to study the PTS under these conditions. Linking computational fluid dynamics analyses to a finite element analysis of stress states and the J integral, provides the input used in the Master Curve approach. Using the Master Curve approach, and under the assumption that only the fracture toughness influence significantly the rupture conditions and the only factor influencing the fracture toughness is the fluence, the probability of fracture under this transient can be approximated. The authors give their outlook about how this approach could be implemented in IAEA and ASME guidelines.

Nomenclature PTS

Pressurized Thermal Shock

NRG

Nuclear Research & Consultancy Group

LOCA

Loss Of Coolant Accident Reactor Pressure Vessel

RPV

ECCS ASME IAEA

Emergency Core Cooling System

American Society of Mechanical engineers International Atomic Energy Agency

SSY BCC

Small Scale Yielding

Body Centred Cubic (crystal structure) Ductile to Brittle Transition Temperature

DBTT

SIF � �

Stress Intensity Factor RPV wall thickness

radius through the cross-section of the RPV

probability for failure

Weibull stress