Session: 04-06: SMRs, Advanced Reactors and Fusion

Paper Number: 130135

130135 - Thermal Analysis and Structural Design of the Main Steam Containment Penetration for Multi-Modular High-Temperature Gas-Cooled Reactor Power Plant 

Abstract: 

The multi-modular high-temperature gas-cooled reactor power plant (MHTGR) is an inherently safe nuclear energy technology for efficient electricity generation and process heat applications. Based on the high-temperature gas-cooled reactor pebble-bed module demonstration plant (HTR-PM), which employed two small modular reactor modules (SMRs) with thermal power of 250 MWth and power generation of approximately 100 MWe per module, several larger versions of the MHTGR with more modules are under design. The designing MHTGR comprises six to ten modules. Each module has a nuclear steam supply system (NSSS) that creates superheated steam and transports the heat through a main steam pipe system. The main steam pipes from different modules run in parallel through the nuclear plant to the conventional plant and connect to one steam turbine.

Unlike the HTR-PM, which includes only one module in each ventilated low-pressure containment (VLPC), the newly designed MHTGR contains all modules in a large circular VLPC for compact design. With this layout, all of the main steam pipes should pass the containment in parallel to the nuclear auxiliary building through a limited area of the containment wall. The steam temperature inside the pipe is over 540 ℃. These high temperature steam pipes are arranged in multiple rows and columns with limited space between each other, which may cause the local overheat of the concrete wall.

The multiple main steam pipe containment penetration design for the MHTGR should meet several essential challenges to keep the concrete wall temperature below the temperature limit. Firstly, the main steam temperature for the MHTGR is one of the highest among all the commercial nuclear plants. The difference in heat transfer temperature between hot pipe and cold concrete wall is enormous. Secondly, the wall surface area between pipes is comparatively limited in comparison to HTR-PM and other nuclear plants. The heat rejection by wall surface is restricted. Thirdly, there is no active cooling on the surface or inside the wall. The cooling source is insufficient. The heat removal only relies on the wall surface heat transfer with the air and the surroundings.

This study first set up a 2-D heat transfer model to discover the influence of the sleeve diameter and insulation thickness on wall temperature. Given the distance between the pipe axial line, there is an optimized sleeve diameter and insulation thickness range since a larger sleeve diameter and insulation thickness decrease the heat transfer from hot pipe to wall but also reduce the wall surface for heat removal. Secondly, the team adopted 3-D modeling to compute and compare the temperature distribution of different potential structural designs. The bellows expansion joint is assembled outside the sleeve instead of inside the sleeve, as done in HTR-PM. Thirdly, the fundamental mechanical parameters were designed considering the temperature distribution and mechanical properties of the materials. A closure head using SA-335 F91 was applied to connect the pipe and the bellows expansion joint and overcome the pipe's high-temperature impact on the bellows expansion joint material and the weld joint. The bellows expansion joint length was calculated considering the pipe movement and the heat conduction from the pipe to the concrete wall through the metal expansion bellows.

Presenting Author: Chaoyi Zhu Institute of Nuclear and New Energy Technology, Tsinghua University

Presenting Author Biography: Dr. Chaoyi Zhu is a research associate professor from the Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing, P.R.China. His research interests include High Temperature Gas-Cooled Reactor heat transport system design, nuclear heat application, waste heat recovery, district heating, nuclear heat desalination, flashing two-phase flow analysis, clean-energy heating, evaporative cooling, and sorption heat pump development.

Authors:

Chaoyi Zhu Institute of Nuclear and New Energy Technology, Tsinghua University
Yiyang Zhang Institute of Nuclear and New Energy Technology, Tsinghua University
Huijie Yan Institute of Nuclear and New Energy Technology, Tsinghua University
Jiyang Fu Institute of Nuclear and New Energy Technology, Tsinghua University
Mei Huang Chinergy Co. Ltd