Title: Some preliminary thermal-hydraulic safety analysis on the NHR-200II reactor

Biography:

Yan Wang is working on Institute of Nuclear and New Energy Technology, China.

Abstract

The NHR-200II as a new 200 MWth nuclear heating reactor, which adopts natural circulation reactor with high passive safety could serve as a safe and economic energy source for combined heat and power of city. The thermal-hydraulic transient response of reactor system during the postulated loss of coolant accident have a significant impact on the containment design and the evaluation on passive safety of reactor. In this research, based on the preliminary design of the NHR-200 II at the current stage, the sub-channel calculation of reactor core and the analysis on some typical loss of coolant accidents were carried on by using COBRA and PCNHR code which was a validated transient analysis programs developed by Tsinghua University. The calculation results show that the NHR-200II has enough safety margin during the normal operation. Under the postulated loss of coolant accident, the reactor core of the NHR-200 II is able to keep being coved by the residual coolant in the reactor pressure vessel during the transient and the relative safety parameters are effectively controlled under the design limitation values without special safety injection like general PWR, which indicates the good safety feature of the NHR-200 II design.