Development of analytical methodology regarding reactor performance and safety characteristics of HTGR; Loss of coolant flow tests

Takamatsu, Kuniyoshi  ; Takeda, Tetsuaki; Nakagawa, Shigeaki  ; Goto, Minoru  

Safety demonstration tests using the HTTR are in progress to verify the inherent safety features and to improve the safety technology and design methodology of high temperature gas-cooled reactors (HTGRs). The numerical analysis code was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel flow channels and twenty temperature coefficients in the core. This paper describes an analytical result of the loss of partial coolant flow test using the newly developed code. The analytical result of transient reactor power shows good agreement with the measured value during the test. Moreover, this paper refers to an analytical result of the loss of coolant flow test. The reactor power decreases to decay heat level due to the negative reactivity feedback effect of the core. Although the reactor power becomes critical again later, the peak power value is very small.