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.
We were studying on the modeling of boiling water reactor fuel assemblies at pin-by-pin level by using Monte Carlo method. The designed boiling water reactor system is cylinder, and the radius of the cylinder is 228 cm. The total active core height is 315.79 cm. The reactor core was divided into the square lattice 7 × 7 type with a constant pitch of 30 cm. The core was surrounded with the reflector. The reflector was surrounded by SS316LN ferritic steel with width of 3 cm. The mixtures 0.2−1% PuF4 and PuO2 were used as fuel. In this study, the effect on the neutronic calculations of PuF4 and PuO2 fuels was investigated in the designed boiling water reactor system. There were calculated k eff , heat deposition and the fission energy in the designed boiling water reactor system. The three-dimensional (3D) modelling of the reactor core and fuel assembly into the designed boiling water reactor system was performed by using MCNPX-2.7.0 Monte Carlo method and the ENDF/B nuclear data library.
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.
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