Stresses in spherical fuel elements of a high-temperature gas-cooled reactor (VTGR) as a result of the heat load and radiation shrinkage of graphite

1984 ◽  
Vol 57 (6) ◽  
pp. 846-850
Author(s):  
V. S. Egorov ◽  
V. S. Eremeev ◽  
E. A. Ivanova
Keyword(s):  
High Temperature ◽  
Heat Load ◽  
Fuel Elements ◽  
High Temperature Gas

Automatic X-ray inspection for escaped coated particles in spherical fuel elements of high temperature gas-cooled reactor

Energy ◽  
2014 ◽  
Vol 68 ◽  
pp. 385-398 ◽  
Author(s):  
Min Yang ◽  
Qi Liu ◽  
Hongsheng Zhao ◽  
Ziqiang Li ◽  
Bing Liu ◽  
...  
Keyword(s):  
High Temperature ◽  
Coated Particles ◽  
X Ray ◽  
Fuel Elements ◽  
High Temperature Gas

Wear behavior of graphitic matrix of fuel elements used in pebble-bed high-temperature gas-cooled reactors against steel

2018 ◽  
Vol 328 ◽  
pp. 353-358 ◽  
Author(s):  
Bin Wu ◽  
Yue Li ◽  
Hong-sheng Zhao ◽  
Shuang Liu ◽  
Bing Liu ◽  
...  
Keyword(s):  
High Temperature ◽  
Wear Behavior ◽  
Pebble Bed ◽  
Fuel Elements ◽  
High Temperature Gas

High Temperature Gas Loop and Experiments of Finned Fuel-Elements

1966 ◽  
Vol 9 (33) ◽  
pp. 166-174
Author(s):  
Yoshizo OKAMOTO ◽  
Shinichi NEGOYA
Keyword(s):  
High Temperature ◽  
Fuel Elements ◽  
Gas Loop ◽  
High Temperature Gas

Gas-Cooled Modular Reactors With Spherical Fuel Elements — Their Inherent Safety and Applications Not Just for Power Generation

2014 ◽  
Author(s):  
Walter Jaeger ◽  
H. J. Hamel ◽  
Heinz Termuehlen
Keyword(s):  
Power Generation ◽  
High Temperature ◽  
Reactor Design ◽  
Inherent Safety ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
Pebble Bed Reactors ◽  
Fuel Elements ◽  
High Temperature Gas

The gas-cooled reactor design with spherical fuel elements, referred to as high-temperature gas-cooled reactors (HTGR or HTR reactors) or pebble bed reactors has been already suggested by Farrington Daniels in the late 1940s; also referred to as Daniels’ pile reactor design. Under Rudolf Schulten the first pebble bed reactor, the 46MWth AVR Juelich reactor (Atom Versuchs-Reactor Jülich) was built in the late 1960s. It was in operation for 22 years and extensive testing confirmed its inherent safety.

A Numerical Analysis on Graphite Dust Deposition and Resuspension in HTR-10 Steam Generator

2014 ◽  
Author(s):  
Wei Peng ◽  
Tian-qi Zhang ◽  
Ya-nan Zhen ◽  
Su-yuan Yu
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Load Capacity ◽  
Fission Products ◽  
Dust Deposition ◽  
Preliminary Calculation ◽  
Transfer Tube ◽  
High Load Capacity ◽  
Fuel Elements ◽  
High Temperature Gas

The behavior of graphite dust is important to the safety analysis of High-Temperature Gas-cooled Reactor (HTGR). The fission products released by fuel elements would enter the primary loop and combine with dust, resulting in that the dust has a high load capacity of cesium, strontium, iodine and tritium. It would bring difficulty and inconvenience to the maintenance and repair of steam generator. Therefore, the behavior of graphite dust in the steam generator is essential to the safety of High Temperature Gas-cooled Reactors. The present study focused on the deposition and resuspension of graphite dust in steam generator of HTR by numerical method. The results show that the graphite dust in steam generator deposits on the surface of heat transfer tube through turbulent deposition, thermophoretic deposition, and other depositional mechanisms, of which thermophoretic deposition is the main mechanism for the particles with the diameter of 2.2μm in the present study. The preliminary calculation result shows that about 6760mg/m2 of graphite dust tends to load on the tube surface.

scholarly journals Computational fluid dynamics investigation of thermal-hydraulic characteristics in a simplified pebbled bed modular reactor core using different arrangements of fuel elements

2017 ◽  
Vol 10 (3) ◽  
pp. 119-127 ◽  
Author(s):  
Yuyu Bai ◽  
Nan Gui ◽  
Xingtuan Yang ◽  
Jiyuan Tu ◽  
Shengyao Jiang
Keyword(s):  
Heat Transfer ◽  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
High Temperature ◽  
Reactor Core ◽  
Face Centered Cubic ◽  
Body Centered Cubic ◽  
Face Centered ◽  
Fuel Elements ◽  
High Temperature Gas

High-temperature gas-cooled reactor is a kind of advanced nuclear reactor in which the core is packed with spherical fuel elements. In high-temperature gas-cooled reactors, the operating temperature is higher than that in ordinary light water reactors. In an attempt to analyze the flow pattern and heat transfer situation to provide reference for the safe operation of the pebble bed reactors, a segment of simplified high-temperature gas-cooled reactor core is simulated with computational fluid dynamics method. Four kinds of arrangement, including simple cubic, body-centered cubic, face-centered cubic, and a combination structure of body-centered cubic and face-centered cubic, are studied, respectively. Based on the simulation results, higher heat transfer capability and lower pebble temperature are obtained in the case with the most compact arrangement. The drag coefficient ( Cd) for four arrangements with different inlet Reynolds number (Re) is obtained and relationship between Re and Cd is analyzed. In addition, a simulation with a broken fuel element in the body-centered cubic fluid domain has been performed. The results show that the presence of broken fuel may result in uneven flow, which will change the heat transfer condition. So it is better to avoid broken fuel element in a high-temperature gas-cooled reactor.

The Gasification of Graphite Matrix in Pebble Bed Reactors

2009 ◽  
Author(s):  
Xinli Yu ◽  
Suyuan Yu
Keyword(s):  
High Temperature ◽  
Corrosion Rate ◽  
Fuel Element ◽  
Pebble Bed ◽  
The Core ◽  
Graphite Matrix ◽  
Matrix Graphite ◽  
Element Matrix ◽  
Fuel Elements ◽  
High Temperature Gas

This paper mainly deals with the simulations of graphite matrix of the spherical fuel elements by steam in normal operating conditions. The fuel element matrix graphite was firstly simplified to an annular part in the simulations. Then the corrosions to the matrix graphite in 10 MW High Temperature Gas-cooled Reactor (HTR-10) and the High Temperature Gas-cooled Reactor—–Pebble-bed Module (HTR-PM) were investigated respectively. The results showed that the gasification of fuel element matrix graphite was uniform and mainly occurred at the bottom of the core in both of the reactors in the mean residence time of the spherical fuel elements. This was mainly caused by the designed high temperature at the bottom. The total mass gasified in HTR-PM was much greater than the HTR-10, while it did not mean much severer corrosion occurred there. As it is known the core volume of HTR-PM is much larger than the HTR-10, which will result in much greater consumed graphite even for the same corrosion rate. The steam only lost about 1 to 3 percent after flowing through the cores in both reactors for different steam conditions. The corrosion of graphite became worse when the steam concentrations increased in helium coolant. The results also indicated that the corrosion rate of fuel element matrix graphite tended to increase slightly with the prolonging of the service time.

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