The development of the fast reactor (FR) cycle is being advanced to utilize plutonium and transuranium (TRU) in Japan. In the fabrication process, it is considered that a fuel pin spirally wrapped with a thin wire is laid horizontally. Then, cooling air flows vertically from the bottom side into the gap of the pin bundle so as to suppress the temperature increase due to decay heat. From the viewpoint of safety assessment during the fabrication, a thermal hydraulic analysis method plays an important role in investigating the maximum temperature and the temperature distribution of the fuel pins. In the present paper, a subchannel analysis tool has been developed. Using the developed tool, the benchmark analysis of the mocked up experiment has been carried out, as well as the numerical investigation of a multidimensional effect of fuel cladding thermal conductivity on the maximum temperature. It is demonstrated that the multidimensional effect of the cladding thermal conductivity is not negligible in the analysis. A good agreement is achieved in the case of a comparatively large clearance size between the side wall and the pin bundle when one considers a natural convection heat transfer at the outermost boundary with a comparatively low computational cost.
- Cross flow
- Low decontaminated fuel
- Subchannel analysis
- Thermal hydraulics
ASJC Scopus subject areas
- Nuclear and High Energy Physics
- Nuclear Energy and Engineering