Analisi CFD del miscelamento di refrigerante nel vessel di un reattore nucleare VVER-1000

Analisi con il codice CFD ANSYS CFX 10.0 dei fenomeni di miscelamento di refrigerante nel vessel di un reattore nucleare VVER-100

Role of CFD Analysis in Nuclear Reactor Licensing and Design

Computational Fluid Dynamics (CFD) is a well-established industrial design tool for non-nuclear applications, helping to reduce design time scales and to improve processes throughout the engineering world, providing a cost-effective and accurate alternative to scale model testing. Within the Nuclear Reactor Safety (NRS) framework, the traditionally adopted tools for safety analysis evaluation (i.e. integral thermal-hydraulic codes) are not capable of predicting the effect of inherently three-dimensional flow fields and mixing phenomena in complex geometries, therefore the application of is considered to potentially bring real benefits in terms of deeper understanding of involved phenomena and of increased safety. However, CFD tools are considered not yet fully mature to be applied to nuclear safety related problems since further code assessment is still necessary. Nevertheless, the intensive code development and assessment work carried out in recent years and the dramatic increase in computing power are quite promising, and CFD already plays an important role as a support tool for NRS analysis. In this framework, the present thesis provides a contribution to the definition of the possible current role and the future perspectives of the application of CFD tools to NRS problems within both a licensing and a design framework. In particular, the present research activity focused on the implementation of CFD techniques within a best estimate methodology to address the licensing analysis of a Nuclear Power Plant (NPP), namely the analysis of the Double Ended Guillotine Break Loss Of Coolant Accident (DEGB-LOCA or 2A-LOCA) scenario of the Atucha-II NPP (CNA-2), which is included into the Chapter 15 of its Final Safety Analysis Report (FSAR). The adopted methodology implies the coupled application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, together with the evaluation of the related uncertainties. A systematic and integrated application of CFD techniques to NRS analysis for licensing purposes is presented, able to go beyond state-of-the-art approaches in this field of application. The present research is also contributing to the assessment of CFD codes in their application to problems related to nuclear safety and technology

ATUCHA-1 NPP CONTAINMENT VENTING ANALYSIS FOLLOWING SBO AND LBLOCA EVENTS BY GOTHIC CODE

Containment behaviour plays a key role in the safety framework of a Nuclear Power Plant. The GOTHIC thermal hydraulic code has been adopted to evaluate the Atucha-1 NPP containment responses during two postulated accident scenarios, Station Black Out and Large Break Loss of Coolant Accident, while assuming the external cooling of the Reactor Pressure Vessel is carried out during the transients. The Atucha-1 NPP has a containment designed to work at full pressure, constituted by a steel sphere enveloped by a concrete shell, and having an annular gap of air in between. The target of the analysis is the evaluation of the effects caused by the additional production of steam in the reactor cavity as a consequence of the ex-vessel cooling, which could cause an excessive pressurization of the containment, and lead to pressure values above the safety limit. The containment pressure and temperature, the distribution of hydrogen in the containment atmosphere and the water hold-up in the most relevant rooms have been monitored as target variables. Each accident scenario was simulated using two different nodalizations, characterized by a different level of refinement. The "detailed" nodalization is meant to be the most refined nodalization according to the available computational resources; having high fidelity three dimensional details, with a high number cells. While the "coarse" nodalization was developed in order to lower the demand for computational resources without significantly compromising the global scenario response. Both nodalizations are characterized by high complexity in the representation of rooms and their connections, e.g. all doors and blow off panels have been simulated to open with the designed differential pressure logic. Both accident transients, for each type of nodalization, were simulated for 200,000 seconds. Results showed that for the Large Break Loss of Coolant Accident pressure is predicted to reach around 5.25 bar, while the Station Black Out Scenario reaches 4.4 bar