NDLI: Strategy for the Development of a DNB Local Predictive Approach Based on Neptune CFD Software

Content Provider	The American Society of Mechanical Engineers (ASME) Digital Collection
Author	Haynes, Pierre-Antoine Pe´turaud, Pierre Montout, Michae¨l Hervieu, Eric
Copyright Year	2006
Abstract	The NEPTUNE project constitutes the thermal-hydraulics part of a long-term joint development program for the next generation of nuclear reactor simulation tools. This project is being carried through by EDF (Electricite´ de France) and CEA (Commissariat a` l’Energie Atomique), with the co-sponsorship of IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and AREVA NP. NEPTUNE is a multi-phase flow software platform that includes advanced physical models and numerical methods for each simulation scale (CFD, component, system). NEPTUNE also provides new multi-scale and multi-disciplinary coupling functionalities. This new generation of two-phase flow simulation tools aims at meeting major industrial needs. DNB (Departure from Nucleate Boiling) prediction in PWRs is one of the high priority needs, and this paper focuses on its anticipated improvement by means of a so-called “Local Predictive Approach” using the NEPTUNE CFD code. We firstly present the ambitious “Local Predictive Approach” anticipated for a better prediction of DNB, i.e. an approach that intends to result in CHF correlations based on relevant local parameters as provided by the CFD modeling. The associated requirements for the two-phase flow modeling are underlined as well as those for the good level of performance of the NEPTUNE CFD code; hence, the code validation strategy based on different experimental data base types (including separated effect and integral-type tests data) is depicted. Secondly, we present comparisons between low pressure adiabatic bubbly flow experimental data obtained on the DEDALE experiment and the associated numerical simulation results. This study anew shows the high potential of NEPTUNE CFD code, even if, with respect to the aforementioned DNB-related aim, there is still a need for some modeling improvements involving new validation data obtained in thermal-hydraulics conditions representative of PWR ones. Finally, we deal with one of these new experimental data needs and present a scaling method for the design of the associated experimentation devoted to the analysis of the dynamics-related modeling of a bubbly flow in PWR representative conditions.
Sponsorship	Nuclear Engineering Division
Starting Page	373
Ending Page	381
Page Count	9
File Format	PDF
ISBN	0791842452
DOI	10.1115/ICONE14-89678
e-ISBN	0791837831
Volume Number	Volume 4: Computational Fluid Dynamics, Neutronics Methods and Coupled Codes; Student Paper Competition
Conference Proceedings	14th International Conference on Nuclear Engineering
Language	English
Publisher Date	2006-07-17
Publisher Place	Miami, Florida, USA
Access Restriction	Subscribed
Subject Keyword	Heart failure Computational fluid dynamics Computer simulation Multiphase flow Computer software Critical heat flux Modeling Pressure Design Databases Numerical analysis Simulation Bubbly flow Two-phase flow Dynamics (mechanics) Nucleate boiling Thermal hydraulics Nuclear reactors Pressurized water reactors
Content Type	Text
Resource Type	Article

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