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Development of thermal hydraulics analysis code ASFRE for fuel assembly of sodium-cooled fast reactor; Modification of distributed resistance model and validation analysis

Kikuchi, Norihiro ; Ohshima, Hiroyuki; Tanaka, Masaaki ; Hashimoto, Akihiko*

For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.

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