The world's most dynamic and flexible transient test reactor is available for research and demonstration activities. The Transient Reactor Test (TREAT) Facility provides a transformational research platform that links science and engineering capabilities for the advancement of fundamental science and nuclear technology. TREAT will directly couple the response of material systems to complex environments experienced under a variety of operational and off-normal transient events anticipated in nuclear reactors. From scientific discovery to technology development and deployment, TREAT will enable the advancement of nuclear material science and engineering required to optimize operation of the existing light water reactor fleet, as well as the advanced reactors of the future.
TREAT is a highly capable test reactor; its unique design offers real-time monitoring of the fuel or materials behavior under postulated reactor accident conditions. This allows scientists to determine the appropriate safe limits for the fuels and materials in nuclear power reactors. TREAT's simple, self-limiting, air-cooled design can safely accommodate multi-pin test assemblies, fostering the study of fuel melting, metal-liquid reactions, and overheated fuel and coolant interactions, as well as the transient behavior of fuels for high-temperature system applications. It also allows for the detailed monitoring of the specimens during a test via the hodoscope, a system that detects fast neutrons and enables real-time evaluation of the fuel behavior within a test sample.
Idaho National Laboratory (INL) recently released footage of a new experiment that simulates what happens to a nuclear fuel pin when it starts to overheat. The new series of tests will ultimately help researchers better understand the safety limits of nuclear fuel.
INL conducted the experiments at its Transient Reactor Test Facility (TREAT) using a first-of-a-kind device that can detect and study the critical heat flux of a nuclear fuel rod. Critical heat flux is the physical phenomenon that occurs when a fuel rod first begins to overheat and can no longer transfer additional heat to the water. This leads to excessive boiling around the surface of the pin and could potentially cause excessive fuel damage.
A Unique Look
The slow-motion video by INL shows the progression of boiling leading up to the point where critical heat flux is reached, when large quantities of water vapor bubbles touch the surface of the fuel rod. The experiment was conducted outside of the test reactor in a specially-designed water-filled capsule that used an electrically heated fuel pin to simulate the conditions. The entire experiment lasted one second, but provided unique insights into this phenomenon.
“Critical heat flux is an important parameter that regulators use to determine the safety limits of nuclear fuel,” said Dr. Colby Jensen, the Transient Testing Technical Leader. “These experiments will help us better understand fuel behavior and to demonstrate how robust safety features of advanced fuel designs will allow more efficient use of those designs.”