Enhancements to Engineering-scale Reactor Pressure Vessel Fracture Capabilities in Grizzly

Technical Report · Fri Sep 01 04:00:00 EDT 2017

DOI:https://doi.org/10.2172/1473611· OSTI ID:1473611

Spencer, Benjamin W. ^[1]; Hoffman, William M. ^[1]; Jiang, Wen ^[1]

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1. Idaho National Lab. (INL), Idaho Falls, ID (United States)

The Grizzly code is being developed to model the effect of aging in nuclear power plant systems, components, and structures. A significant part of this effort has been to develop capabilites to model the effects of embrittlement in reactor pressure vessels on their integrity. This includes both modeling of microstructure and engineering property evolution, and engineering-scale probabilistic fracture mechanics analysis. This report documents recent advances to the engineering-scale fracture mechanics capability for evaluation of RPV integrity under transient loading in Grizzly. These developments are in three areas: probabilistic fracture mechanics, general reduced order models for fracture a flaw locations, and improvments to the XFEM capability used in Grizzly for fracture mechanics analysis. The combination of these developments brings Grizzly closer to a state where it can be applied in production as a general tool for engineering-scale fracture mechanics analysis. The probabilistic fracture mechanics capabilities are included in the 1.5 testing version of Grizzly, and will be further refined in preparation for production use in the 2.0 released planned for fiscal year 2018.

• Research Organization: Idaho National Laboratory (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1473611
• Report Number(s): INL/EXT--17-43427-Rev000
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
97 MATHEMATICS AND COMPUTING
Embrittlement
Fracture Mechanics
Reactor Pressure Vessel