Title: Inherent safety and ATWS analysis of High Temperature Gas-cooled Reactor

Biography:

Yanhua Zheng has her expertise in nuclear reactor thermal-hydraulic design and accident analysis. She is responsible for the thermal-hydraulic design and accident analysis of the Chinese 200 MWe HTR-PM project. She has comprehensive and in-depth study on typical accidents and key phenomena of the pebble-bed HTGR, especially on the water-ingress accident, air-ingress accident, uncertainty analysis and so on. Her research covers the software development and V&V, accident management procedure development, and BDBA mitigation method design of HTGR. She is also the leading researcher or main participant of several National Science & Technology Major Projects and National Natural Science Foundation Projects of China.

Abstract

The modular high temperature gas-cooled reactor (HTGR), recognized as a candidate for the Generation IV nuclear energy system technology, has well-known inherent safety features. A commercial-scale 200 MWe Pebble-bed Modular High Temperature gas-cooled Reactor (HTR-PM) has been designed and constructed in Shandong Province, China. Most of the construction and installation work have been finished and the connection to the electric grid will be expected in 2019 or the first half of 2020.In this paper, the design and the inherent safety feature of the HTR-PM has been introduced. Several Anticipated Transient Without Scram (ATWS) accidents, a type of Beyond Design Basis Accident (BDBA) receiving high attention especially in Pressurized Water Reactor (PWR) analysis, have been studied, including the reactivity introduction ATWS, loss of off-site power ATWS, depressurized loss of coolant ATWS.Calculation results prove that, even in such kind of BDBAs with very low probability, the inherent safety design of HTR-PM can guarantee the reactor shut-down itself by negative temperature feedback. During the accidents, the decay heat of the reactor can be transferred to the environment safely by heat conduction, natural convection and radiation, and the fuel temperature and the reactor pressure vessel (RPV) temperature would never exceed the limitation. The large release of the fission products would not happen.