Abstract The objective of the present work is to apply a BE code to determine an important safety margin of a certain type of research reactor that is Onset of Flow Instability (OFI). These BE code systems have been developed for power reactors and were extensively validated against experimental data relevant to this kind of reactors and their application to research reactors is not straight full. During the last years, these BE codes have been enhanced and were suggested to be used for research reactors applications. Thus, the objective of the present work is to investigate the RELAP5/Mod 3 code system capabilities in predicting the onset of flow instability in uniformly-heated vertical fuel plates under research reactor’s operating conditions (low pressure-low temperatures). For this purpose, the Oak Ridge National Laboratory Thermal Hydraulic Test Loop (THTL) has been chosen to be reproduced. This thermalhydraulic loop has been designed and built to reflect operating conditions of the Advanced Neutron Source Reactor at ORNL. The results obtained during this framework have shown some limitations of RELAP5/Mod 3.2 in the critical region where OFI is expected and the closure relationship between mass flux and heat flux is not verified by the recalculated data
Analysis of Onset of Flow Instability in Rectangular Heated Channel by Relap5/3.2 Code System
D'AURIA, FRANCESCO SAVERIO
2005-01-01
Abstract
Abstract The objective of the present work is to apply a BE code to determine an important safety margin of a certain type of research reactor that is Onset of Flow Instability (OFI). These BE code systems have been developed for power reactors and were extensively validated against experimental data relevant to this kind of reactors and their application to research reactors is not straight full. During the last years, these BE codes have been enhanced and were suggested to be used for research reactors applications. Thus, the objective of the present work is to investigate the RELAP5/Mod 3 code system capabilities in predicting the onset of flow instability in uniformly-heated vertical fuel plates under research reactor’s operating conditions (low pressure-low temperatures). For this purpose, the Oak Ridge National Laboratory Thermal Hydraulic Test Loop (THTL) has been chosen to be reproduced. This thermalhydraulic loop has been designed and built to reflect operating conditions of the Advanced Neutron Source Reactor at ORNL. The results obtained during this framework have shown some limitations of RELAP5/Mod 3.2 in the critical region where OFI is expected and the closure relationship between mass flux and heat flux is not verified by the recalculated dataI documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
