Neighborhood min distance descriptor for kinship verification

Multimedia Tools and Applications - The verification of parental and family relationships based on the facial appearance of subjects is a recent topic that attracted the attention of the computer...     

Thermal-hydraulic behavior of the NUR nuclear research reactor during a fast loss of flow transient

Nuclear safety analysis remains of crucial importance for both the design and the operation of nuclear reactors. Safety analysis usually entails the simulation of several selected postulated accidents, which can be divided into two main categories, namely reactivity insertion accident (RIA) and loss of flow accident (LOFA). In this paper, thermal-hydraulic simulations of fast LOFA accident were carried out on the new core configuration of the material test research reactor NUR. For this purpose, the nuclear reactor analysis PARET code was used to determine the reactor performance by calculating the reactor power, the reactivity and the temperatures of different components (fuel, clad and coolant) as a function of time. It was observed that during the transient the maximum clad temperature remained well below the critical temperature limit of 110 °C, and the maximum coolant temperature did not exceed the onset of nucleate boiling point of 120 °C. It is concluded that the reactor can be operated at full power level with sufficient safety margins with regard to such kind of transients.     

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT

This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria).

The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor.

The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior.

3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.     

NUR research reactor safety analysis study for long time natural convection (NC) operation mode

The current work represents a part of an overall effort being undertaken to improve NUR nuclear research reactor neutron utilization capabilities using a new core configuration. A RELAP5 model for the reactor NUR under natural convection (NC) operating mode has been developed. The model represents internal pool reactor components with the corresponding geometry, point neutron kinetics, and thermal hydraulics characteristics. An experimental device was designed and implemented in the reactor pool for monitoring the inlet and outlet core temperatures, and other pool temperature positions during NC-operating mode.In this paper, unprotected fast reactivity insertion and total flow blockage of the flapper valve transients have been investigated under NC operating conditions.The achieved steady-state results were found to be in good qualitative agreement with measurements. The results obtained from the transient (FRIA) study were compared to similar approach from recent literature. The second transient herein considered, is an attempt to predict the reactor core thermal-hydraulic behavior under a total flow blockage of the (NC) flapper valve. The latter could be considered as useful contribution for updating the safety analysis report.