Technical and QA plan: Boiling behavior during flow instability

Technical Report · Fri Sep 06 04:00:00 EDT 1991

DOI:https://doi.org/10.2172/10129393· OSTI ID:10129393

Coutts, D A

The coolant flow in a nuclear reactor core under normal operating conditions is kept as a subcooled liquid. This coolant is evenly distributed throughout the multiple flow channels with a uniform pressure profile across each coolant flow channel. If the coolant flow is reduced, the flow through individual channels will also decrease. A decrease in coolant flow will result in higher coolant temperatures if the heat flux is not reduced. When flow is significantly decreased, localized boiling may occur. This localized boiling can restrict coolant flow and the ability to transfer heat out of the reactor system. The maximum operating power for the reactor may be limited by how the coolant system reacts to a flow instability. One of the methods to assure safe operation during a reducing flow transient, is to operate at a power level below that necessary to initiate a flow excursion. Several correlations have been used to predict the conditions which will proceed a flow excursion. These correlations rely on the steady state behavior of the coolant and are based on steady-state testing. There are two significant points which this project will try to identify. The first is when vapor first forms on the channel surface. This might be designated as the Nucleate Vapor Transition. (Steady state equivalent is ONB). The second is when the vapor formation rate is large enough to lead to flow instability and thermal excursion. This point might be designated as the Significant Vapor Transition. (Steady state equivalent is OSV). A correlation will be developed to relate established steady state relations with the behavior of transient systems.

• Research Organization: Westinghouse Savannah River Co., Aiken, SC (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC09-89SR18035
• OSTI ID: 10129393
• Report Number(s): WSRC-MS--91-514; ON: DE92009093
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
420400
FLOW RATE
FLUID FLOW
HEAT TRANSFER
HEAT TRANSFER AND FLUID FLOW
HYDRAULICS
LOSS OF COOLANT
NUCLEATE BOILING
QUALITY ASSURANCE
REACTOR COOLING SYSTEMS
SAVANNAH RIVER PLANT
STABILITY
SUBCOOLED BOILING