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The 100kWth Compact High-Temperature Reactor (CHTR) is envisaged as a technology demonstrator for the Indian high-temperature reactor programme enabling process heat applications of nuclear energy such as thermochemical splitting of water for hydrogen. Coupled neutronics–thermal-hydraulics analyses of safety transients in prototype reactors like CHTR are essential for safe design and operation under the high-temperature regime. In this regard, the unprotected loss of flow accident (ULOFA) of CHTR core in full-power configuration has been investigated with the indigenously developed and validated code system. The ULOFA of CHTR has also been analysed for severe transient case considering the simultaneous accidental withdrawal of one of the rods for power regulation. The simulations of these transients have been carried out with indigenously developed point kinetics code PATH. The code is integrated with the thermal-hydraulics module to predict the temperature distribution in the fuel assembly for feedback. The variations of nuclear power, core reactivity and radial and axial temperature profile in an average-powered and peak-powered channel of CHTR are predicted for the 90 min of transients after the trigger. The simulation methodology and results of the safety transients of CHTR are reported in this article.
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