Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal–hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal–Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics–thermal–hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.
Application of coupled code technique to a safety analysis of a standard MTR research reactor
D'AURIA, FRANCESCO SAVERIO
2009-01-01
Abstract
Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal–hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal–Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics–thermal–hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
