Victor Hugo Sánchez Espinoza

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

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.

Uncertainty and Sensitivity Studies with TRACE-SUSA and TRACE-DAKOTA by Means of Steady State BFBT Data

The subject of the present paper is the uncertainty and sensitivity studies for steady state BFBT results including pressure drop and void fraction measurements. The investigations are performed with TRACE (version 5.0 patch 2), thermal hydraulic modeling, and SUSA and DAKOTA; both tools are for the evaluation of uncertainties and sensitivities. For this purpose, the NUPEC BFBT experimental data base is used. The advantage of applying two different uncertainty and sensitivity tools in combination with TRACE is that the user effect can be excluded. Since in both cases the TRACE model of the BFBT bundle is identical the differences in the results are related to the capabilities of the uncertainty and sensitivity tools. The reference results with TRACE show that the code is very well able to represent both single- and two-phase flows even though it is a 1D coarse mesh system code. For selected cases, an uncertainty study was performed. Even though a reduced number of uncertain parameters are considered in the DAKOTA investigation, compared to the one with SUSA, similar results are obtained. The results indicate also that even small parameter variations can yield to rather large variations of the selected output parameters.

Uncertainty and Sensitivity Study in the Frame of TRACE Validation for Reflood Experiment

Abstract The validation of computer codes related to the thermal-hydraulic analyses of nuclear reactors is a challenging undertaking because of the complexity of the phenomena involved, e.g., boiling, condensation, and mixing. In the frame of the ongoing validation of the best-estimate system code TRACE, the present paper focuses on the phenomena taking place during the quenching of the hot surface of the fuel rod simulator with cold water. Since TRACE describes the physical phenomena with empirical correlations derived from experiments, it is necessary to ensure that these correlations are valid if applied to similar experiments but different boundary conditions. By means of an uncertainty and sensitivity study, the influence of the empirical models and their associated uncertainties on selected output parameters is quantified and the parameters with the largest sensitivity are evaluated.

Thermal-Hydraulic Evaluation of an LBE Cooled 19-Pin Bundle in the Frame of TRACE Validation

In the frame of the validation and verification of the thermal hydraulic system code TRACE for lead alloy cooled systems, a 19 pin bundle is analyzed, in particular, the influence of the spacers. TRACE is updated and improved in order to be used for the analyses of liquid metal cooled reactors. For the present investigation, isothermal hydraulic experiments, carried out at the Karlsruhe Lead Laboratory, are used to validate the pressure losses related to wall friction and spacers. In this experimental set-up, a 19 rod bundle, cooled by liquid lead-bismuth eutectics (LBE) is used. In addition, thermal analyses are performed in order to evaluate the axial temperature profile of coolant and cladding and to validate the chosen heat transfer models. Since the thermal experiments have not been performed yet, the validation is done by comparing TRACE results to CFD results following a unique approach for the turbulent Prandtl number. Special emphasize is paid to the validation of the models to account for heat transfer enhancement at and in the vicinity of spacers due to increased turbulences.Copyright

Post-Test Investigations of BFBT Critical Power Tests With Trace

The validation of heat transfer models of safety analysis codes such as TRACE is very important due to the strong interaction of the thermal hydraulics parameters with the core neutronics. TRACE is the reference system code of the US NRC for LWR. It is being developed and extensively validated within the international CAMP-program. In this paper, the validation of heat transfer models of TRACE related to the prediction of the critical power is presented. The validation is based on a large number of critical power tests performed in the NUPEC BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility in Japan. These tests were analysed with the TRACE Version 5 RC 2. The comparison of predictions with the experimental data shows good agreement. The developed TRACE model and the comparison of experimental data with code results will be presented and discussed.Copyright

A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR). By applying codes like CFD (computational fluid dynamics) and SP3 (simplified transport) reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3) based neutron kinetics (NK) code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

On the Applicability of Pool Boiling Models in TRACE for the Evaluation of a Counter-Current LBE-Water Heat Exchanger

A counter current LBE-Water heat exchanger is investigated with the system code TRACE with emphasize on the water side. Due to the liquid metal coolant on the outer side of the heat exchanger the present boundary conditions for the water side (pressure, mass flux, heat flux, etc.) might be out of range of normal LWR application and need therefore additional validation. The review of the physical models of TRACE, mainly the ones for pool boiling, indicated that these models might not be applicable to the present heat exchanger design. Alternative pool boiling models have been implemented into TRACE, namely the ones of Forster-Zuber and of Labuntsov. The comparison shows that the heat transfer coefficient of the original model is about 30% higher than with the alternative models but with a wall heat flux which is almost identical for the three models. This is caused by the much lower heat transfer coefficient on the LBE side which dictates the global heat transfer. The resulting wall temperature and the power which can be transferred from the hot LBE side to the water are lower for the alternative models. That indicates that with the standard model too optimistic results might be calculated.Copyright