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COBRA IIIc/MIT-2: a digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements. Final report

Technical Report ·
OSTI ID:5059627
;
The COBRA IIIc/MIT-2 computer program computes the flow and enthalpy in rod-bundle nuclear fuel element subchannels during both steady state and transient conditions. It uses a mathematical model which considers both turbulent and diversion crossflow mixing between adjacent subchannels. Each subchannel is assumed to contain one-dimensional, two-phase, separated, slip-flow. The two-phase flow structure is assumed to be fine enough to define the void fraction as a function of enthalpy, flow-rate, heat-flux, pressure, position and time. At the present time, steady-state two-phase flow correlations are assumed to apply to transients. The mathematical model neglects sonic velocity propagation; therefore, it is limited to transients where the transient times are greater than the time for a sonic wave to pass through the channel. The equations of the mathematical model are solved by using a semi-explicit finite difference scheme. This scheme also gives a boundary-value flow solution for both steady state and transients where the boundary conditions are the inlet enthalpy, inlet mass velocity, and exit pressure.
Research Organization:
Massachusetts Inst. of Tech., Cambridge (USA). Energy Lab.
OSTI ID:
5059627
Report Number(s):
PB-82-180233; MIT-EL-81-018
Country of Publication:
United States
Language:
English

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Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
C CODES
COMPUTER CODES
ENTHALPY
FLUID FLOW
FLUID MECHANICS
FUEL ASSEMBLIES
FUEL CHANNELS
FUEL ELEMENT CLUSTERS
HEAT FLUX
HYDRAULICS
MATHEMATICAL MODELS
MECHANICS
PHYSICAL PROPERTIES
REACTOR ACCIDENTS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY
SAFETY
STEADY-STATE CONDITIONS
THERMAL ANALYSIS
THERMODYNAMIC PROPERTIES
TRANSIENTS
TWO-PHASE FLOW