• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.295-306
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• 2009

The subchannel grade void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility were evaluated with the subchannel analysis code MATRA and the system code MARS. Fifteen test series from five different test bundles were selected for an analysis of the steady-state subchannel void distributions. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5% to 25%. The results of the transient calculations were also similar and were highly feasible. However, the computational aspects of the two codes were clearly different.

• Nuclear Engineering and Technology
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• v.10 no.2
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• pp.79-86
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• 1978

The object of this study was to obtain data on air-distributions in two-phase up flow in vertical rod-bundle test-section. The test-section in this study was a hexagonal shaped 61-rod bundle where each rod was wrapped with helical spacers. The variables were flow rates of air and water and air inlet positions. Experimental data were obtained at the outlet of the test-section. The experiments were performed in two parts. Firstly, data were taken at increasing flow rates of air keeping water flow rates constant, and secondly, at simultaneous increase of air and water flow rates. At each flow condition, air supply position could be changed to 4 different positions. Data obtained by electrical void-needle technique were analyed and are presented here in graphical forms for comparison. The results of this study demonstrate qualitatively that air-distribution tends to be more uniform as water flow rates are increased. The air supply positions have noticeable effects on the pattern of air-distribution.

• Nuclear Engineering and Technology
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• v.56 no.12
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• pp.5106-5117
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• 2024

Mixing Vane Grid (MVG) is considered as one of the most important components in the fuel assembly which not only plays the role of supporting the rod bundles but also improves the Critical Heat Flux (CHF) in the reactor core. Modeling and measuring the flow behavior accurately in the rod bundle is the key to understanding and learning complex grid performance in the fuel assembly and will develop high performance MVG. Usually, the fuel assembly in the reactor core consists of 17 × 17 or 16 × 16 rod bundles, it is hardly to use the original MVGs to perform study. The representative smaller prototypical grids are applied. Different bundle sizes are used including 1 × 1, 2 × 1, 3 × 3 and 5 × 5 et al. It is an absolute question of how the smaller size rod bundles are prototypical that could fully reflect the true flow and heat transfer behavior in a reactor core. In this paper, the effect of bundle size on flow and heat transfer is investigated under sizes of 2 × 1, 3 × 3 and 5 × 5. Firstly, the boundary settings in 2 × 1 are studied and the surface averaged secondary flow and local flow at the gap with 5 × 5 results are compared. Then the 3 × 3 and 5 × 5 bundle sizes are compared under subcooled flow. The center subchannels temperature and the void fraction distributions are analyzed. The effect of non-prototypical cold walls on heat transfer is discussed. The study shows that, different bundle sizes will produce different flow phenomena in the rod bundle, the flow pattern may not be the same with the reactor core fuel assembly, the typical bundle size selection should be based on the research purpose.