Approximate model for evaluation of thermal-hydraulic transients associated with rapid power increase in research nuclear reactor

• Main Authors: Lipka Maciej, Madejowski Gawel, Prokopowicz Rafal, Pytelt Krzysztof
• Format: Article
• Language: English
• Published: VINCA Institute of Nuclear Sciences 2020-01-01
• Series: Nuclear Technology and Radiation Protection
• Subjects: reactor safety; thermal hydraulics; lumped-parameter model; heat transfer; research reactor
• Online Access: http://www.doiserbia.nb.rs/img/doi/1451-3994/2020/1451-39942004310L.pdf

A simple model, for the estimation of changes in the nuclear fuel element cladding temperature as well as the conditions of the formation of the onset of nucleate boiling, is proposed. The results of this estimation are sufficient to assess the effect of the transient with the peak of the reactor's power on the device's safety, without the need for time-consuming thermal calculations. The basic parameters determined using the proposed model are the maximum wall temperature of the device in a hot spot, the time constant of the wall temperature change, and the course of changes in the onset of nucleate boiling ratio in time. The model may be used for investigating the thermal behavior of thin heat-generating and water-cooled elements (such as fuel elements or uranium irradiation targets) during rapid power rise. The results of the temperature estimation with the proposed model were tested considering the hot spot in the MR-6 type nuclear fuel. The SN code with coupled kinetics and thermal-hydraulics, developed in the MARIA reactor was used to validate the results. The maximum cladding temperature, the thermal time constant and the onset of nucleate boiling ratio parameter simulated by the SN code and the proposed scheme appeared to be very consistent.