Containment pressure assessment using the COBRA-NC computer code

Conference · Thu Jan 01 04:00:00 EST 1987 · Trans. Am. Nucl. Soc.; (United States)

OSTI ID:6808286

Gangadharan, A; Hassan, Y A

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-
• Conference Information: Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 55
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BLOWDOWN
BWR TYPE REACTORS
C CODES
COMPUTER CODES
CONTAINMENT
CONTAINMENT SYSTEMS
DESIGN
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
MECHANICS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR CORE DISRUPTION
REACTOR SAFETY
REACTORS
RELIABILITY
SAFETY
THREE-DIMENSIONAL CALCULATIONS
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS