An Experimental Study on Insulating Property of Helium in High Temperature Gas Cooled Reactor

2014 ◽  
Author(s):  
Shan Yue ◽  
Xingnan Liu ◽  
Zhengang Shi
Keyword(s):  
High Temperature ◽  
Electrical Equipment ◽  
Structure Design ◽  
Nuclear Industry ◽  
Parallel Plane ◽  
Primary Coolant ◽  
Experimental Platform ◽  
Insulating Properties ◽  
High Temperature Gas ◽  
Better Than

HTGR, short for high temperature gas cooled reactor, has gained a lot of attention in nuclear industry. Gas helium, 7MPa in pressure, is used as primary coolant of HTR-PM in where there are a lot of electrical equipment. Insulating property of helium is worse than that of air according to Paschen curves and there are very few articles or related standards about insulating property of high pressure gas helium, which makes the electrical equipment structure design lack of basis. In this study, an experimental platform for testing insulating performance is designed, based on which the experiments of testing the withstanding voltages of penetration assemblies and the breakdown voltages of parallel plane electrodes at different pressures are carried out. The results show that for both the penetration assemblies and the parallel plane, their breakdown voltages in helium are far lower than in air under the same condition of 15°C /0.1MPa. For the penetration assemblies, their insulating properties in helium at 150°C/7MPa are better than those in air at 15°C/0.1MPa.

scholarly journals The Design of Turbomachinery for the Gas Turbine (Direct Cycle) High Temperature Gas Cooled Reactor Power Plant

1977 ◽  
Author(s):  
R. G. Adams ◽  
F. H. Boenig
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Gas Turbine ◽  
Power Conversion ◽  
Reactor Power ◽  
Working Fluid ◽  
Primary Coolant ◽  
Reactor Power Plant ◽  
Direct Cycle ◽  
High Temperature Gas

The Gas Turbine HTGR, or “Direct Cycle” High-Temperature Gas-Cooled, Reactor power plant, uses a closed-cycle gas turbine directly in the primary coolant circuit of a helium-cooled high-temperature nuclear reactor. Previous papers have described configuration studies leading to the selection of reactor and power conversion loop layout, and the considerations affecting the design of the components of the power conversion loop. This paper discusses briefly the effects of the helium working fluid and the reactor cooling loop environment on the design requirements of the direct-cycle turbomachinery and describes the mechanical arrangement of a typical turbomachine for this application. The aerodynamic design is outlined, and the mechanical design is described in some detail, with particular emphasis on the bearings and seals for the turbomachine.

Conceptual Structure Design of High Temperature Isolation Valve for High Temperature Gas Cooled Reactor

2010 ◽  
Author(s):  
Shoji Takada ◽  
Kenji Abe ◽  
Yoshiyuki Inagaki
Keyword(s):  
High Temperature ◽  
Hydrogen Production ◽  
Production System ◽  
Conceptual Structure ◽  
Structure Design ◽  
Radioactive Material ◽  
Normal Operation ◽  
Temperature History ◽  
Allowable Limit ◽  
High Temperature Gas

The high temperature isolation valve (HTIV) is a key component to assure the safety of a high temperature gas cooled reactor (HTGR) connected with a hydrogen production system, that is, protection of radioactive material release from the reactor to the hydrogen production system and combustible gas ingress to the reactor at the accident of fracture of an intermediate heat exchanger and the chemical reactor. The HTIV used in the helium condition over 900 °C, however, has not been made for practical use yet. The conceptual structure design of an angle type HTIV was carried out. A seat made of Hasteloy-XR is welded inside a valve box. Internal thermal insulation is employed around the seat and a liner because high temperature helium gas over 900 °C flows inside the valve. Inner diameter of the top of seat was set 445 mm based on fabrication experiences of valve makers. A draft overall structure was proposed based on the diameter of seat. The numerical analysis was carried out to estimate temperature distribution and stress of metallic components by using a three-dimensional finite element method code. Numerical results showed that the temperature of the seat was simply decreased from the top around 900 °C to the root, and the thermal stress locally increased at the root of the seat which was connected with the valve box. The stress was lowered below the allowable limit 120 MPa by decreasing thickness of the connecting part and increasing the temperature of valve box to around 350 °C. The stress also increased at the top of the seat. Creep analysis was also carried out to estimate a creep-fatigue damage based on the temperature history of the normal operation and the depressurization accident.

scholarly journals Management of Radioactive Waste Containing Graphite: Overview of Methods

Energies ◽  
2020 ◽  
Vol 13 (18) ◽  
pp. 4638
Author(s):  
Leon Fuks ◽  
Irena Herdzik-Koniecko ◽  
Katarzyna Kiegiel ◽  
Grazyna Zakrzewska-Koltuniewicz
Keyword(s):  
High Temperature ◽  
Radioactive Waste ◽  
Nuclear Power ◽  
Information Service ◽  
Nuclear Industry ◽  
Current Status ◽  
Triso Fuel ◽  
Irradiated Graphite ◽  
High Temperature Gas

Since the beginning of the nuclear industry, graphite has been widely used as a moderator and reflector of neutrons in nuclear power reactors. Some reactors are relatively old and have already been shut down. As a result, a large amount of irradiated graphite has been generated. Although several thousand papers in the International Nuclear Information Service (INIS) database have discussed the management of radioactive waste containing graphite, knowledge of this problem is not common. The aim of the paper is to present the current status of the methods used in different countries to manage graphite-containing radioactive waste. Attention has been paid to the methods of handling spent TRISO fuel after its discharge from high-temperature gas-cooled reactors (HTGR) reactors.

A Comprehensive Study of the 14C Source Term in the 10 MW High-Temperature Gas-Cooled Reactor

Radiocarbon ◽  
2019 ◽  
Vol 61 (5) ◽  
pp. 1169-1183 ◽  
Author(s):  
X Liu ◽  
W Peng ◽  
L Wei ◽  
M Lou ◽  
F Xie ◽  
...  
Keyword(s):  
High Temperature ◽  
Power Plants ◽  
Activity Concentration ◽  
Specific Activity ◽  
Theoretical Calculations ◽  
Total Porosity ◽  
Sensitivity Study ◽  
Human Beings ◽  
Primary Coolant ◽  
High Temperature Gas

ABSTRACTWhile assessing the environmental impact of nuclear power plants, researchers have focused their attention on radiocarbon (14C) owing to its high mobility in the environment and important radiological impact on human beings. The 10 MW high-temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China that adopted helium as primary coolant and graphite spheres containing tristructural-isotropic (TRISO) coated particles as fuel elements. A series of experiments on the 14C source terms in HTR-10 was conducted: (1) measurement of the specific activity and distribution of typical nuclides in the irradiated graphite spheres from the core, (2) measurement of the activity concentration of 14C in the primary coolant, and (3) measurement of the amount of 14C discharged in the effluent from the stack. All experimental data on 14C available for HTR-10 were summarized and analyzed using theoretical calculations. A sensitivity study on the total porosity, open porosity, and percentage of closed pores that became open after irradiating the matrix graphite was performed to illustrate their effects on the activity concentration of 14C in the primary coolant and activity amount of 14C in various deduction routes.

Steam-Turbine-Driven Circulators for High-Temperature Gas-Cooled Reactors—Part II: Development

1974 ◽  
Vol 96 (2) ◽  
pp. 102-108
Author(s):  
V. J. Barbat ◽  
D. Kapich ◽  
F. C. Dahms ◽  
J. Yampolsky
Keyword(s):  
High Temperature ◽  
Steam Turbine ◽  
Prestressed Concrete ◽  
Axial Compressor ◽  
Development Program ◽  
Single Stage ◽  
Coolant Fluid ◽  
Primary Coolant ◽  
Experimental Development ◽  
High Temperature Gas

The High-Temperature Gas-Cooled reactor is characterized by integration of the primary coolant circuit and components within a Prestressed Concrete Pressure Vessel. This concept requires particular features and assurance in the circulators that are used to circulate the primary coolant fluid. Each circulator employs a single-stage axial compressor driven by a single-stage steam turbine which is in series with the main steam turbo-generator. The circulator is lubricated by water and is capable of variable speed operation. The conception and design of the circulator were discussed in the first part of this paper. The present part describes the experimental development program of the series steam-turbine-driven helium circulator, which verified and supported the design methods and demonstrated the operational capability of the circulator in all the possible modes that could occur in plant operation.

Conceptual Structure Design of High Temperature Isolation Valve for High Temperature Gas Cooled Reactor

2011 ◽  
Vol 133 (11) ◽  
Author(s):  
Shoji Takada ◽  
Kenji Abe ◽  
Yoshiyuki Inagaki
Keyword(s):  
High Temperature ◽  
Hydrogen Production ◽  
Production System ◽  
Three Dimensional ◽  
Chemical Reactor ◽  
Conceptual Structure ◽  
Structure Design ◽  
Radioactive Material ◽  
Allowable Limit ◽  
High Temperature Gas

The high temperature isolation valve (HTIV) is a key component to assure the safety of a high temperature gas cooled reactor connected with a hydrogen production system for protections of radioactive material release from the reactor to the hydrogen production system as well as of combustible gas ingress to the reactor at the accident of fracture of an intermediate heat exchanger and the chemical reactor. However, the HTIV has not been made for practical use in the helium condition over 900°C yet. The conceptual structure design of an angle type HTIV was carried out. A seat made of Hasteloy-XR is welded inside a valve box. Internal thermal insulation is employed around the seat and a liner because the high temperature helium gas flows inside the valve. The inner diameter of the top of seat was set 445 mm based on fabrication experiences of valve makers. A draft overall structure was proposed based on the diameter of the seat. The numerical analysis was carried out to estimate the temperature distribution and stress of metallic components by using a three-dimensional finite element method code. Numerical results showed that the temperature of the seat was simply decreased from the top around 900°C to the root, and the thermal stress locally increased at the root of the seat, which was connected with the valve box. The stress was lowered below the allowable limit 120 MPa by decreasing the thickness of the connecting part and increasing the temperature of the valve box to around 350°C. The stress also increased at the top of the seat. Creep analysis revealed that the intactness of the HTIV is kept after the assumed operation cycles of the plant life as well as at the depressurization accident.

Application Research of Memory Alloy Washer on Power Equipment Safety and Energy Saving

2013 ◽  
Vol 842 ◽  
pp. 719-724
Author(s):  
Tian Xiang Chen ◽  
Xiao Lin Fang ◽  
Hong Tian
Keyword(s):  
High Temperature ◽  
Contact Resistance ◽  
Electrical Equipment ◽  
Initial Contact ◽  
Power Equipment ◽  
Application Research ◽  
Remarkable Effect ◽  
Experimental Platform ◽  
Equipment Safety ◽  
Reducing Energy

Contact resistance is an essential cause of electrical equipment joint heating. An experimental platform was built to measure and compare temperature and contact resistance of memory alloy spring washers and ordinary spring washers when heated by taking the same heavy current, on condition that both of the counterparts had the same initial contact resistance and temperature. Results showed that memory alloy washer were more helpful to limit the joint temperature rise by reducing contact resistance from their features of tempting to restore to original shape when heated to a high temperature. Therefore, the memory alloy washers were proved to have remarkable effect on ensuring safety of electric power equipments and reducing energy loss.

Steam Turbine Driven Circulators for High-Temperature Gas-Cooled Reactors—Part I: Design

1974 ◽  
Vol 96 (2) ◽  
pp. 95-101 ◽  
Author(s):  
J. Yampolsky ◽  
L. Cavallaro ◽  
G. C. Thurston ◽  
M. K. Nichols
Keyword(s):  
High Temperature ◽  
Steam Turbine ◽  
Prestressed Concrete ◽  
Axial Compressor ◽  
Development Program ◽  
Single Stage ◽  
Coolant Fluid ◽  
Primary Coolant ◽  
Experimental Development ◽  
High Temperature Gas

The High-Temperature Gas-Cooled reactor is characterized by integration of the primary coolant circuit and components within a Prestressed Concrete Pressure Vessel. This concept requires particular features and assurance in the circulators that are used to circulate the primary coolant fluid. Each circulator employs a single-stage axial compressor driven by a single-stage turbine which is in series with the main steam turbogenerator. The circulator is lubricated by water and is capable of variable speed operation. An extensive design and development program was carried out to provide a family of circulators for a range of reactor sizes. This paper considers the design features of the series steam turbine driven helium circulator and the basis for the adopted design solutions. Part II of this paper considers the experimental development program.

Study on the Break Accidents of the HTR-PM Primary Loop

2016 ◽  
Author(s):  
Minggang Lang ◽  
Ximing Sun ◽  
Yanhua Zheng
Keyword(s):  
High Temperature ◽  
Pressure Vessel ◽  
Relief Valve ◽  
Primary Coolant ◽  
Primary Loop ◽  
The Core ◽  
Reactor Building ◽  
Temperature And Pressure ◽  
Coolant System ◽  
High Temperature Gas

In thermal hydraulics designing and safety analysis of the High Temperature gas-cooled Reactor-Pebble Bed (HTR-PM), the THERMIX code was used to study the behavior of helium in the primary coolant system. Once the helium leaks out of the primary loop through a break on the pressure boundary or an inadvertent open relief valve, it is difficult to simulate the conditions of the room where the release occurred with THERMIX. In this paper, the latest version of RELAP5/MOD4 was used to simulate the behavior of the helium released to the containment rooms. A RELAP5/MOD4 input deck of the HTR-PM, consisting of the core, the primary coolant system, the secondary loop and the containment, was developed and evaluated in this paper. Based on the model, this paper simulated the accidents consequences of large breaks or small breaks near the inlet or the outlet of the helium circulator located inside the steam generator pressure vessel. The calculating results illustrate that the temperature of the helium flowing into the reactor building through the break was no more than 280°C even after an un-isolating large break. The analysis shows that the systems function to scram the reactor and to monitor the core temperature and pressure after accidents would not be affected by breaks.

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