Validation of NEPTUNE-CFD Two-Phase Flow Models Using Experimental Data

Abstract

This paper deals with the validation of the two-phase flow models of the CFD code NEPTUNEC-CFD using experimental data provided by the OECD BWR BFBT and PSBT Benchmark. Since the two-phase models of CFD codes are extensively being improved, the validation is a key step for the acceptability of such codes. The validation work is performed in the frame of the European NURISP Project and it was focused on the steady state and transient void fraction tests. The influence of different NEPTUNE-CFD model parameters on the void fraction prediction is investigated and discussed in detail. Due to the coupling of heat conduction solver SYRTHES with NEPTUNE-CFD, the description of the coupled fluid dynamics and heat transfer between the fuel rod and the fluid is improved significantly. The averaged void fraction predicted by NEPTUNE-CFD for selected PSBT and BFBT tests is in good agreement with the experimental data. Finally, areas for future improvements of the NEPTUNE-CFD code were identified, too.

Comparison between compressible, dilatable and incompressible fluid hypotheses efficiency in liquid conditions at high pressure and large temperature differences

“…Trio_CFD is an open source code [19] and more details on the code developments are integrally accessible on the web site [8]. Neptune_CFD is not an open source code, but a lot of papers about the code validation are accessible in literature [18,29] and more information can be found on the web site [28].…”

A homogeneous relaxation low mach number model

“…By considering the instantaneous limits of each relaxation process, we obtain a hierarchy of models, each with partial equilibrium in one or more of the aforementioned variables. Most industrial codes within the nuclear community -for instance CATHARE [11], FLICA [25], NEPTUNE_CFD [45], RELAP [9], THYC [37], TRAC [51], WAHA [27] -rely on these models.…”

The multiscale thermal‐hydraulic simulation for nuclear reactors: A classification of the coupling approaches and a review of the coupled codes

“…Nevertheless, new methods known as the computational multifluid dynamics (CMFD), which dedicates to simulate the two‐phase flow on mesoscale or microscale are already one of the most promising trends for reactor applications . Recent research devoted to enhance and validate the CFD codes for two‐phase flow problems of reactor safety has advanced …”

“…37 Recent research devoted to enhance and validate the CFD codes for two-phase flow problems of reactor safety has advanced. [38][39][40] Considering the trends and the current capabilities of the different thermal-hydraulic codes for design and safety evaluations, the development of multiscale thermal-hydraulic simulations has been initiated in different teams worldwide, in order to enhance the nuclear reactor thermal-hydraulic simulation. Normally, the phenomena occurring on system scale, component scale, and mesoscale are simulated by system code, subchannel code, and CFD code, respectively.…”