Reactor Core Behavior of NUR Research Reactor during a Protected Fast Loss of Flow Accident
DOI:
https://doi.org/10.38208/ete.v5.1175Keywords:
Transient Conditions, Thermal hydraulic, Code system Relap5, Nuclear research reactor, Safety analysis, Loss of flow accidentAbstract
The best estimate thermal hydraulic codes are the most used code for thermal hydraulic and safety analysis of nuclear reactors, regarding their ability to provide a more realistic nuclear reactor behavior during normal and transient operating conditions. Moreover, the use of such codes is essential in order to ensure safe and reliable operation of nuclear reactors. In this study, for safety assessment purposes, the code system Relap5/mod3.2 was used to predict the thermal-hydraulic behavior of the NUR research reactor during a protected fast loss of flow accident. This was accomplished through the reactor core modeling by using a one-hot channel and an average one as well to present the remaining channels in the reactor core. Then, the transient variation of fuel, clad and coolant temperatures in both channels, further to, the safety criteria: - The critical heat flux ratio (CHFR) and – The onset of Flow Instability Ratio (OFIR) were evaluated and analyzed from safety point of view, where it was concluded that the occurrence of the protected fast loss of flow accident in NUR research reactor does not lead to any safety issue or damage to the fuel. This justified by the fact that, the clad peak values are 73.2°C and 55.3°C respectively for the hot and average channels in their first peak while they become equal to 72.7°C and 62.0°C in their second peak which are far below the limit value (450°C) during transients, in the other hand, the lower values of CHFR are 4.6 and 9.9 respectively for the hot and average channels in their first minimum point while they become equal to 3.4 and 7.7 in their second minimum point which does not fall below the imposed safety limit 1.5 during transients.
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