Session: 04-07: SMRs, Advanced Reactors and Fusion

Paper Number: 133432

133432 - Development of Thermal–hydraulic and Safety Analysis Code for a Heat Pipe Cooled Reactor 

Abstract: 

Heat pipe cooled reactors use the natural circulation flow of two-phase working medium in alkali metal heat pipe to export the heat generated by nuclear fission to the energy conversion system. Heat pipe cooled reactors have gained attention as a potential solution for nuclear power generation in space and deep-sea applications due to their uncomplicated design, scalability, safety, and dependability. While various models of heat pipe-cooled reactors have already been proposed, there remains a lack of a rapid and efficient thermal-hydraulic code for assessing the operational characteristics and safety analysis of these reactors. In this paper, a transient thermal-hydraulic analysis code was developed to evaluate the transient characteristics of the core of heat pipe-cooled reactor.

The code comprises three integral modules: a point reactor neutron physics module, a heat pipe module, and a core heat transfer module. The heat pipe module, developed using the thermal resistance method, has been validated with experimental data. The core heat transfer module employs a multi-channel model and considers air gap heat transfer and heat conduction between individual channels. The results of the computational fluid dynamics (CFD) model and the developed heat transfer module for the core were compared under the same boundary conditions. Both modules demonstrated high agreement with reference values.

A 5MW heat pipe-cooled reactor, designed by Argonne National Laboratory (ANL), served as the subject of this analysis. The code efficiently calculated both steady-state contributions and variable load operating condition. In steady-state calculations, the results indicated that the temperature distribution characteristics accurately mirrored the operating behavior of the heat pipe cooled reactor. Under variable load operating condition, the reactor is able to safely complete power conversion and maintain stability at different power levels. Furthermore, the code was used to analysis the operational characteristics of typical accidents, including reactivity-induced accidents, heat pipe failure, and loss of heat sink accidents. During the heat pipe failure accident, the heat from the failed fuel assembly could be transferred through adjacent heat pipes. In the reactivity-induced accident, the reactor power stabilized to a new steady state after a period. In the accident of a loss of heat sink, the reactor core safely shut down due to negative feedback, with core temperature initially spiking then decreasing. Throughout all transient processes of these accidents, the fuel pin temperatures maintained a considerable safety margin, underscoring the reactor's robust safety performance under various accident scenarios. This code provides an effective tool for the design and safety analysis of heat pipe cooled reactors.

Presenting Author: Guanghui Jiao Harbin Enginering University

Presenting Author Biography: Guanghui jiao , a Phd student of Harbin Engineering Univerasity, currently engaged in the safety analysis of heat pipe cooled reactors.

Authors:

Guanghui Jiao Harbin Enginering University
Genglei Xia Harbin Engineering University
Tao Zhou Harbin Engineering University
Jianjun Wang Harbin Enginering UNiversity