Silicon Carbide Clad Graphite Matrix Fuel Elements

2009 ◽  
pp. 318-318-6
Author(s):  
KM Taylor ◽  
CH McMurtry
Author(s):  
Haitao Wang ◽  
Xin Wang

Spherical fuel elements with a diameter of 60mm are basic units of the nuclear fuel for the pebble-bed high temperature gas-cooled reactor (HTR). Each fuel element is treated as a graphite matrix containing around 10,000 randomly distributed fuel particles. The essential safety concept of the pebble-bed HTR is based on the objective that maximum temperature of the fuel particles does not exceed the design value. In this paper, a microstructure-based boundary element model is proposed for the large-scale thermal analysis of a spherical fuel element. This model presents detailed structural information of a large number of coated fuel particles dispersed in a spherical graphite matrix in order that temperature distributions at the level of fuel particles can be evaluated. The model is meshed with boundary elements in conjunction with the fast multipole method (FMM) in order that such large-scale computation is performed only in a personal desktop computer. Taking advantage of the fact that fuel particles are of the same shape, a similar sub-domain approach is used to establish the temperature translation mechanism between various layers of each fuel particle and to simplify the associated boundary element formulation. The numerical results demonstrate large-scale capacity of the proposed method for the multi-level temperature evaluation of the pebble-bed HTR fuel elements.


1957 ◽  
Vol 1 ◽  
pp. 387-398 ◽  
Author(s):  
D. S. Flikkema ◽  
R. V. Schablaske

AbstractIt has been found possible to determine quickly the concentrations of molybdenum and ruthenium in non-radioactive alloys representative of high burn-up reactor fuels by the method of X-ray emission spectrometry. Preliminary steps of chemical dissolution and separation are not required. The alloys, essentially ternaries of molybdenum and ruthenium with uranium, are being studied because they are considered to typify the alloys which will result from cycling uranium fuel elements through the sequence of fabrication, use and pyro.metallurgical processing.The analytical procedure involves sampling of the ingot by slicing with a silicon carbide wheel at the plane of interest and reducing the surface to the flatness and finish obtained by a five-minute grinding and polishing operation. In the X-ray spectrograph the flat surface is examined for the intensities of its molybdenum and ruthenium K emission lines, with counting times of one to eight minutes. Calibration plots of intensity versus chemically determined weight per cent are established and used for subsequent sets of analyses.


Author(s):  
Milan F. Hrovat ◽  
Karl-H. Grosse ◽  
Richard Seemann

The molded block fuel element (FE) also called monolith is a molded body, consisting of a substantially isotropic highly crystalline graphite matrix, fuel regions within the same matrix and cooling channels. The fuel regions contain the fuel in the form of coated particles which are well bonded to the remaining graphite matrix, so that both parts of the block form a monolithic structure. The monolith meets the requirements for the very high temperature reactors attaining helium outlet temperatures above 1000°C. To fabricate the molded blocks FE demonstration plant was erected and put into operation. The equipment worked without malfunction. The produced block FEs meet the specifications of GA machined block FEs. All specimens and block segments irradiated at temperature up to 1600°C and max. fast fluence E > 0, 1 MeV of 11×1021 n/cm2 show perfect behaviour without any damage.


2017 ◽  
Vol 2017 ◽  
pp. 1-6 ◽  
Author(s):  
Xiaotong Chen ◽  
Zhenming Lu ◽  
Hongsheng Zhao ◽  
Bing Liu ◽  
Junguo Zhu ◽  
...  

For High-Temperature Gas-Cooled Reactor in China, fuel particles are bonded into spherical fuel elements by a carbonaceous matrix. For the study of fuel failure mechanism from individual fuel particles, an electrochemical deconsolidation apparatus was developed in this study to separate the particles from the carbonaceous matrix by disintegrating the matrix into fine graphite powder. The deconsolidated graphite powder and free particles were characterized by elemental analysis, X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectrometer (EDS), and ceramography. The results showed that the morphology, size distribution, and element content of deconsolidated graphite matrix and free particles were notably affected by electric current intensity. The electrochemical deconsolidation mechanism of spherical fuel element was also discussed.


Author(s):  
T. T. Hlatshwayo ◽  
N. G. van der Berg ◽  
E. Friedland ◽  
J. B. Malherbe ◽  
P. Chakraborty

In a modern high-temperature nuclear reactor, safety is achieved by encapsulating the fuel elements by CVD-layers of pyrolytic carbon and silicon carbide (SiC) to prevent the fission products release. Some studies have raised doubts on the effectiveness of SiC layer as a diffusion barrier to fission fragments due to 110mAg released from the coated particle at high temperatures ranging from 1500°C to 1600°C [1].


Author(s):  
Xinli Yu ◽  
Suyuan Yu

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.


Author(s):  
Jinhua Wang ◽  
Bing Wang ◽  
Bin Wu ◽  
Yue Li ◽  
Haitao Wang

With the continuous development of the nuclear power technology in the world, all countries in the world are becoming more and more interested in the inherent safety of nuclear power technology, while the research and development of the spherical bed type high temperature gas cooled reactor nuclear power technology in China has formally catered to this demand. As a major national science and technology project, since the construction of the high temperature gas cooled reactor demonstration project (HTR-PM) since 2012, the civil construction of the nuclear island has been basically completed, the installation of equipment has been carried out orderly, and many process systems have entered debugging and operation stage gradually. As an important auxiliary process system, fuel handling and storage system for online refueling of the pebble bed high temperature gas cooled reactor, plays an important role in relation to the stable operation of the reactor. The main functions of the fuel handling and storage system are loading the fresh fuel elements and unloading the spent fuel elements which has reached its target burnup continuously for reactor operation, the spent fuel elements would be discharged into the spent fuel canister firstly, when the spent fuel storage canister is full of spent fuel, the canister would be sealed through welding method, and then the spent fuel canister would be transferred and stored in the spent fuel storage silo with the ground crane system. The fuel element of the pebble bed high temperature gas cooled reactor is spherical fuel element with graphite matrix, the fuel elements will have friction and collision with the inner wall of the pipeline in transporting process, which will produce graphite dust, the graphite dust should be removed continuously though filtration method, so as not to affect the fuel elements transportation in pipeline. This article focus on the production mechanism and filtering method of the graphite dust in graphite matrix fuel element transporting process in pipeline, to study the graphite dust removal technology, and then we could provide theoretical guidance for the design and operation of the key system and equipment for HTR-PM.


1988 ◽  
Vol 64 (2) ◽  
pp. 107-112
Author(s):  
A. S. Chernikov ◽  
T. A. Mireev ◽  
V. V. Teslenko ◽  
S. D. Kurbakov ◽  
L. I. Mikhailichenko ◽  
...  

2019 ◽  
Vol 525 ◽  
pp. 1-6
Author(s):  
Chi Zhang ◽  
Xiaotong Chen ◽  
Bing Liu ◽  
Zengtong Jiao ◽  
Luhao Fan ◽  
...  

Sign in / Sign up

Export Citation Format

Share Document