Containment pressure assessment using the COBRA-NC computer code
The prediction of pressure behavior of a multicompartment containment to a water blowdown transient has recently gained much attention, mainly due to the nuclear power plant accident at Chernobyl. This study simulates and compares the data from a pressurized water blowdown experiment to approximate the accident condition assumed for design of full-pressure containments using the three-dimensional COBRA-NC thermal-hydraulic computer code. The experiment is a standard problem of the Committee on the Safety of Nuclear Installations, which was performed in the model containment of Battelle Institut, Frankfurt, FRG, within the framework of the German Reactor Safety Research Program. Different heat transfer correlations were investigated. The Uchida correlation was found to best simulate the experimental data. The COBRA-NC code is a thermal-hydraulic code for transient analysis of nuclear reactor components of light water reactors including the reactor core, reactor vessel, steam generators, and the reactor containment building. It provides a two-component, two-fluid, three-field representation of two-phase flow. Three momentum, four mass, and two energy equations are solved for the fluid. The simulation test was modeled using the lumped parameter option.
- Research Organization:
- Texas A and M Univ., College Station (USA)
- OSTI ID:
- 6808286
- Report Number(s):
- CONF-8711195-; TRN: 88-033617
- Journal Information:
- Trans. Am. Nucl. Soc.; (United States), Vol. 55; Conference: American Nuclear Society winter meeting, Los Angeles, CA, USA, 15 Nov 1987
- Country of Publication:
- United States
- Language:
- English
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