An analysis of air-water flow phenomena due to a pipe break under sub-atmospheric pressures using TRACE

Abstract

Abstract For advanced research reactor designs that have a high core power with downward coolant flow in the core, it is important to analyze a unique postulated pipe break scenario under sub-atmospheric pressures. In order to benchmark the capability of thermal–hydraulic codes such as TRACE to predict flow behavior under these conditions, we must first compare TRACE analyses to reference experimental data for air–water flow near atmospheric pressure and temperature conditions. Given a successful comparison between predictions and test data, we then use TRACE to analyze air–water two-phase flow phenomena during a postulated pipe break accident with the initial conditions prototypic of a particular research reactor design. The normalized flow rate, pressure, and void fractions were predicted to describe the overall behavior of the pipe break phenomena. These results indicate that the amount of air ingested from the environment into the reactor coolant system through the pipe break has the dominant effect on the downstream void fraction as well as flow and pressure response, and thereby the time available (delay time) to take compensatory actions before deleterious effects are felt by the reactor cooling system. This defined delay time is the figure of merit that can be used for design of the system. Since the two-phase flow under sub-atmospheric pressure in the system can be affected by various design conditions, as well as initial and boundary conditions, the sensitivity analyses were performed to better understand their effect. A set of parameters were chosen based on their quantitative effect on the overall transient behavior; i.e., pool inventory, system design flow rate, and pipe break size. The various quantitative characteristics of the sensitivity analysis results showed that TRACE was capable of analyzing this unique air–water two-phase flow phenomena under sub-atmospheric pressure and will be utilized in the safety analysis for this research reactor system.