Coupling of Neutron Transport and Probabilistic Fracture Mechanics Codes for Analysis of Embrittled Reactor Pressure Vessels

Abstract

Abstract In light water nuclear reactors, the reactor pressure vessel (RPV) plays the critical role of containing the reactor and coolant and is expected to maintain its integrity under a variety of conditions. The RPV is subjected to high neutron flux and temperature, which lead to material embrittlement and increase its susceptibility to fracture. The degree of embrittlement is highly dependent on the local fluence, so it is critical to accurately characterize the spatial distribution of fluence in the RPV when performing a probabilistic fracture mechanics (PFM) analysis, which individually evaluates the flaws introduced in the manufacturing process. Radiation transport codes typically focus on computing the neutron distribution within the nuclear reactor core to understand phenomena relevant to reactor physics. However, recent developments in the Virtual Environment for Reactor Applications (VERA) allow it to solve for a fuel-pin resolved in-core fission source and use Monte Carlo methods to compute neutron fluence accumulation at locations away from the core, such as within the RPV, with unprecedented accuracy. Grizzly, a PFM code, can read in a 3D map of the fluence computed by VERA, and directly use that in the PFM analysis. This represents a significant advance in the accuracy of the fluence used in PFM calculations. This paper provides an overview of this modeling approach demonstrates its application on PFM analyses of representative RPVs.