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Numerical simulation of bubble behavior in PWR rod bundle by interface tracking method

界面追跡法によるPWR燃料集合体内気泡流挙動解析

吉田 啓之 ; 小野 綾子 

Yoshida, Hiroyuki; Ono, Ayako

Critical heat flux (CHF) is one of key parameter to design a fuel bundle of nuclear reactors. Currently, CHF is evaluated based on experimental database performed by large scale test facilities. It is difficult to perform experiments by large scale test facility, because of requirement of huge cost and long time. Therefore, the design of new fuel bundle is difficult, and this is one of important issues related to thermal-hydraulics of nuclear reactors. To resolve this issue, it is considered that the CFD is one of the important tools. If we can simulate thermal-hydraulic phenomena in detail, CHF may be evaluated based on simulated phenomena. However, performing numerical simulation of thermal-hydraulic phenomena in detail is difficult, because absence of physical model related to CHF and numerical simulation method to perform two-phase flow simulations in rod bundles. In JAEA, we started a research project to construct an evaluation method of CHF based on the multiphase CFD technique. In the first step of this project, we performed numerical simulations of bubble behavior in PWR rod bundles by using TPFIT. TPFIT is a numerical simulation code based on an interface tracking method developed in JAEA. In this numerical simulation, we performed single phase flow simulations in PWR rod bundles by using STAR-CCM$$_{+}$$. These results were used as an inlet boundary condition of numerical simulation of bubble behavior in the PWR rod bundle. In the numerical simulations of STAR-CCM$$_{+}$$, existences of grid spacer and vane were calculation parameters. In the results, numerical simulation of bubbly flow in PWR rod bundle under high pressure and temperature conditions. It is confirmed that the bubbly flow behavior is affected by the existence of spacer grid.

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