Influence of the 3-D Phenomena on the Safety Parameters during a ULOF Accident in the MYRRHA Reactor

TY - GEN

T1 - Influence of the 3-D Phenomena on the Safety Parameters during a ULOF Accident in the MYRRHA Reactor

AU - Andreoli, Francesco

AU - Bisogni, Dario

AU - Belloni, Francesco

AU - Scheveneels, Guy

N1 - Score=3

PY - 2015/8

Y1 - 2015/8

N2 - In a pool type liquid metal cooled reactor like MYRRHA, the influence of 3-D thermal-hydraulic phenomena during Loss Of Flow (LOF) transients can potentially have an important impact on safety-relevant parameters. 3-D temperature distributions and local pressure gradients may affect the evolution of the coolant mass flow during the transition from forced to natural convection, with the possible generation of flow instabilities and dissipating flows. Furthermore, the presence of stagnant volumes may influence the characteristic propagation time of perturbations through the system. The 1-D computational system codes used to perform reactor safety analyses were originally developed for loop type reactor designs that foresee the coolant flowing in pipes, with energy losses mainly due to wall friction. Therefore, this class of codes is not validated to simulate correctly the physics of the phenomena occurring in a pool type reactor. The objective of this work is to assess the shortcomings of 1-D system codes in predicting the response of the MYRRHA reactor to a LOF event, identifying the 3-D safety-relevant phenomena that can have an influence on the transient evolution. The adopted strategy is comparing the transient simulation results of the RELAP5 thermal-hydraulic system code with reference CFD simulations.

AB - In a pool type liquid metal cooled reactor like MYRRHA, the influence of 3-D thermal-hydraulic phenomena during Loss Of Flow (LOF) transients can potentially have an important impact on safety-relevant parameters. 3-D temperature distributions and local pressure gradients may affect the evolution of the coolant mass flow during the transition from forced to natural convection, with the possible generation of flow instabilities and dissipating flows. Furthermore, the presence of stagnant volumes may influence the characteristic propagation time of perturbations through the system. The 1-D computational system codes used to perform reactor safety analyses were originally developed for loop type reactor designs that foresee the coolant flowing in pipes, with energy losses mainly due to wall friction. Therefore, this class of codes is not validated to simulate correctly the physics of the phenomena occurring in a pool type reactor. The objective of this work is to assess the shortcomings of 1-D system codes in predicting the response of the MYRRHA reactor to a LOF event, identifying the 3-D safety-relevant phenomena that can have an influence on the transient evolution. The adopted strategy is comparing the transient simulation results of the RELAP5 thermal-hydraulic system code with reference CFD simulations.

KW - MYRRHA

KW - Loss of flow

KW - CFD

KW - system code

KW - 3-D phenomena

UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/ezp_141240

UR - http://knowledgecentre.sckcen.be/so2/bibref/13140

M3 - In-proceedings paper

T3 - NURETH - International Topical Meeting on Nuclear Reactor Thermalhydraulics

SP - 4703

EP - 4716

BT - 16th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16)

CY - United States

T2 - NURETH 2015- 16th International Topical Meeting on Nuclear Reactor Thermalhydraulics

Y2 - 30 August 2015 through 4 September 2015

ER -