A long-term flow-induced vibration and wear test was performed for a full-scale 17×17 PWR fuel mockup, and the test results were compared with numerical simulations. The flow-induced vibration on a fuel assembly or fuel rods may cause Grid-to-Rod Fretting (GTRF) and result in the leakage of fuel rods in PWRs. GTRF involves non-linear vibration of a fuel rod due to the excitation force induced by coolant flow around a fuel rod. So, the numerical simulation is performed by VITRAN (Vibration Transient Analysis Non-linear) and Computational Fluid Dynamics (CFD). VITRAN code was developed by Westinghouse to simulate fuel rod flow induced vibration and GTRF. In this paper, it was confirmed that the code can reproduce GTRF wear for NFI fuel assembly. CFD calculation is performed to obtain the axial and lateral flow velocity around the fuel rods, reflecting detailed geometries of fuel assembly components like bottom nozzle, spacer grids. The numerical simulation reasonably reproduced the vibration and wear test for NFI fuel assembly.

A new code coupled between a sub-channel analysis code and a computational multi-fluid dynamics (CMFD) code was applied to a PWR rod bundle with mixing vane grid. The code was developed to predict the departure from nucleate boiling (DNB). This is a new technology of CMFD based on abundant experience and models developed for sub-channel analysis: CMFD computed void distribution to fit the average value calculated by two-phase models in sub-channel analysis code. A new source term represented centripetal motion of small bubbles in the wake behind a rising vapor slug. In order to apply the code to a PWR rod bundle, effects of local mass flux, pressure and mixing vane grid was modeled. The present method was applied to the analysis for DNB tests of simulated PWR fuel assembly with mixing vane grids. The result showed that, using only a critical void fraction, reasonable prediction was achieved in a wide range of flow condition and a variety of flow regimes.

Accurate analyses of turbulence structure and void fraction distribution are quite important in designing and safety evaluation of various industrial equipments using gas-liquid two-phase flow such as nuclear reactor, etc. Using turbulence model of two-phase flow and models of bubble behaviors in bubble flow and slug flow, systematic analyses of distributions of void fraction, averaged velocity and turbulent velocity were carried out and compared with experimental data. In bubbly flow, diffusion of bubble and lift force are dominant in determining void fraction distribution. On the other hand, in slug flow, large scale turbulence eddies which convey bubbles into the center of flow passage are important in determining void fraction distribution. In turbulence model, one equation turbulence model is used with turbulence generation and turbulence dissipation due to bubbles. Mixing length due to bubble is also modeled. Using these bubble behavior models and turbulence models, systematic predictions were carried out for void distributions and turbulence distributions for wide range of flow conditions of two phase flow including bubbly and slug flow. The results of predictions were compared with experimental data in round straight tube with successful agreement. In particular, concave void distributions in bubbly flow and convex distribution in slug flow were well predicted based on the present model.

To investigate the effect of mixing-vane shape, heat flux at departure from nucleate boiling (DNB) and pressure loss were measured. Computational fluid dynamics (CFD) was utilized to discuss the flow control. The pressure loss and the DNB tests were performed in a water and a Freon loops, respectively. Two mixing-vanes were designed to have same projection area but different inclination. The rod-bundle was 5 by 5 and 17 by 17 respectively at the water and Freon tests. The experimental results showed that the slightly inclined mixing-vane produced the same DNB heat flux as the deeply inclined mixing-vane and did smaller pressure loss than it. Pressure loss of the two mixing-vane grids was different in spite of the same projection area. The result of CFD showed a swirl flow decaying along the main stream in the axial direction. The swirl was stronger in the deeply inclined mixing-vane, however it decayed faster whereas one maintained long in the slightly inclined mixing-vane. This result suggested that the deep inclination caused a steep change in axial momentum to induce strong turbulence diffusion. This flow structure did not change the DNB heat flux because the two-phase discontinuity dominated the phenomena. This study provided a successful example of flow control in a mixing-vane grid.

A computational multi-fluid dynamics (CMFD) code that predicted the void distribution in sub-channel was coupled with sub-channel analysis code to predict departure from nucleate boiling (DNB). The main assumption was that the void fraction near heated wall was the dominant parameter in DNB. A sub-channel analysis code was used to calculate three dimensional distribution of sub-channel averaged values of mass flux, void fraction, density and quality. These were used as a boundary condition in the CMFD code to predict local void fraction in a subchannel. A bubble diffusion equation was used assuming the wall peak void distribution caused by turbulence. The present method was applied to the analysis of DNB tests. The coupled codes showed a reasonable profile of void fraction in a rod bundle and reproduced DNB heat flux at low void fraction. To investigate this analysis result, the local condition were compared with a DNB flow regime map. This suggested an approach to improve the predictability: the critical void fraction should be modified at low void fraction condition; the bubble diffusion model should be modified to handle the flow regime transition from the isolated nucleation type to churn turbulent flow type.

Numerical simulation code for predicting void distribution in two-phase turbulent flow in a sub-channel was developed. The purpose is to obtain a profile of void distribution in the sub-channel. The result will be used for predicting a heat flux at departure from nucleate boiling (DNB) in a rod bundle for the pressurized water reactor (PWR). The fundamental equations were represented by a generalized transport equation, and the transport equation was transformed onto the generalized coordinate system fitted to the rod surface and the symmetric lines in the sub-channel. Using the finite-volume method the transport equation was discretized for the SIMPLE algorism. The flow field and void fraction at the steady state were calculated for different average void fractions. The computational result for atmospheric pressure condition was successfully compared with experimental data. Sensitivity analysis for the PWR condition was performed, and the result showed that the secondary flow slightly contributed to homogenizing the void distribution.