**9. Acknowledgment**

Financial supports from the NSERC/NRCan/AECL Generation IV Energy Technologies Program and NSERC Discovery Grant are gratefully acknowledged.

#### **10. References**


Since the development of SCWRs is still in the conceptual design stage, it is worth to further investigate heat transfer and neutronic aspects of high and enhanced thermal-conductivity fuels. In regards to high thermal-conductivity and enhanced thermal-conductivity fuels, this study recommends the use of UC and UO2-BeO, respectively. This use is conditional on the assurance of chemical compatibility, mechanical behavior, and irradiation behavior of these fuels under the SCWR conditions. In addition, the development of new fuel bundle designs, which will comply with the design temperature limits on the fuel and the sheath, is necessary. New fuel-bundle designs, which would result in lower fuel centerline

Heat transfer at supercritical conditions has been studied by many researchers; however, still there is a need to improve the correlations used to predict the heat transfer coefficient. To the knowledge of the authors, none of the available heat-transfer correlations predicts the deteriorated heat transfer regime. The lack of capability to predict such phenomenon may result in melting of the sheath. Thus, it is significantly important to develop either look-up tables or heat transfer correlations that would predict the deteriorated heat transfer regime.

Financial supports from the NSERC/NRCan/AECL Generation IV Energy Technologies

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**8. Conclusion** 

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**8** 

*Japan* 

**Development of an Analytical Method** 

**Assemblies Under Earthquake Condition** 

Safe operation of nuclear reactors under earthquake conditions cannot be guaranteed because the behavior of thermal fluids under such conditions is not yet known. For instance, the behavior of gas-liquid two-phase flow during earthquakes is unknown. In particular, fluctuation in the void fraction is an important consideration for the safe operation of a nuclear reactor, especially for a boiling water reactor (BWR). The void fraction in the coolant is one of the physical parameters important in determining the thermal power of the reactor core, and fluctuations in the void fraction are expected to affect the power of the plant.

To evaluate fluctuation in the void fraction, numerical simulation is the most effective and realistic approach. In this study, we have developed a numerical simulation technique to predict boiling two-phase flow behavior, including fluctuation in the void fraction, in a fuel

In developing this simulation technique, we selected a three-dimensional two-fluid model as an analytical method to simulate boiling two-phase flow in a fuel assembly because this model can calculate the three-dimensional time variation in boiling two-phase flow in a large-scale channel such as a fuel assembly while incurring only a realistic computational cost. In addition, this model has been used to successfully predict the void fraction for a steady-state boiling two-phase flow simulation (Misawa, et al., 2008). We expect that the development of the boiling two-phase flow analysis method for a fuel assembly under earthquake conditions can be achieved by improving the three-dimensional two-fluid model analysis code ACE-3D (Ohnuki, et al., 2001; Misawa, et al., 2008), which has been developed

This paper describes an analytical method for boiling two-phase flow in a fuel assembly under earthquake conditions by improving ACE-3D and shows how the three-dimensional behavior of boiling two-phase flow under these conditions is evaluated by the improved

**1. Introduction** 

assembly under earthquake conditions.

by the Japan Atomic Energy Agency.

ACE-3D.

**on Water-Vapor Boiling Two-Phase** 

**Flow Characteristics in BWR Fuel** 

Takeharu Misawa, Hiroyuki Yoshida

and Kazuyuki Takase *Japan Atomic Energy Agency* 


