Abstract
Fuel rod cladding failure in a nuclear reactor produces different phenomena related to vibrations and fluid–structure interaction. The most significant aspect of those phenomena is the creation of a pressure wave at the failure position and its propagation in the coolant fluid flowing around the fuel rod. An accurate understanding of the propagation of the pressure wave around the fuel rod can help us design a method to detect a failure, determine its position, and estimate some of its characteristics with a single and simple sensor, such as a pressure sensor or a piezoelectric acoustic sensor, that can be mounted relatively far from the failure. Such a method can be useful for the monitoring of nuclear fuel rods, where instrumentation possibilities are restricted (because of neutron flux, radiation, high temperature, and available space) as well as for any kind of application involving annular ducts and limited instrumentation possibilities. The current paper is related to the specific application of nuclear fuel rod monitoring. It deals with preliminary numerical simulations that are necessary to know the evolution of a fluid pressure profile along the system containing the rod. They are carried out by finite element methods, using the EUROPLEXUS code. They provide the necessary information about the propagation of pressure waves around the rod to design measurement and signal processing methods as well as properly interpret experimental results from tests in industrial reactors, research reactors, or experimental mock-ups. They also provide some information that could not be experimentally obtained because of the constraints in a nuclear environment. Despite the specific application we show in this article, similar calculation methods, theoretical observations, and results interpretations can be easily adapted to the other mentioned applications.
Highlights
Introduction iationsIn most pressurized water nuclear reactors (PWR), fuel material consists of small pellets that are inserted in thin metallic tubes
When numerical results are compared to experimental results from a failure test in a real reactor, the failure length in the model should be compared to the size of the reaction area where the over-pressure is produced in the real test and to the pressure profile in this area
( xi + x j − c f ∆t) with xi and x j representing the axial positions of the two points from which pressure histories are extracted, ∆t representing the time difference of arrival (TDOA) between the two points, and c f representing the primary wave speed
Summary
In most pressurized water nuclear reactors (PWR), fuel material consists of small pellets that are inserted in thin metallic tubes. Such a metallic tube is called a cladding, and the element consisting of the cladding filled with fuel pellets is referred to as a fuel rod. The cladding is supposed to physically separate the fuel material from the surrounding water. The failure can produce a pressure wave that propagates in the surrounding water. Such a phenomenon is more likely to occur during a Reactivity-Initiated
Sign-in/Register to view highlights & summary Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
