scholarly journals Evaluation of Oxidation Performance of TRISO Fuel Particles for Postulated Air-Ingress Accident of HTGR

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Fangcheng Cao ◽  
De Zhang ◽  
Qingjie Chen ◽  
Hao Li ◽  
Hongqing Wang
Keyword(s):  
Reaction Rates ◽  
Good Oxidation Resistance ◽  
Triso Fuel ◽  
Rate Determining Step ◽  
Parabolic Law ◽  
Oxidation Performance ◽  
Triso Particles ◽  
Fuel Particles ◽  
Air Ingress ◽  
High Temperature Gas

In a high-temperature gas-cooled reactor, the integrity of tristructural-isotropic-(TRISO-) coated fuel particles ensures the safety of the reactor, especially in case of an air-ingress accident. The oxidation of TRISO particles with the outer layers of silicon carbide (SiC) was performed at temperatures of 900°C–1400°C in air environment. Both the microstructure and phase composition of the SiC layers were studied. The results showed that the SiC layers had a good oxidation resistance below 1100°C. However, the amorphous silica on the SiC layers formed at 1200°C and gradually crystallized at 1400°C with the presence of microcracks. The reaction rates of the SiC layers were determined by measuring the silica thickness. It was proposed that the oxidation of the SiC layers followed the linear-parabolic law with the activation energy of 146 ± 5 kJ/mol. The rate-determining step of the oxidation was the diffusion of oxygen in silica.

A HR-SEM Investigation of the Microstructure of PBMR Test Fuel Particles

2008 ◽  
Author(s):  
N. G. van der Berg ◽  
J. B. Malherbe ◽  
A. J. Botha
Keyword(s):  
High Temperature ◽  
South African ◽  
Nuclear Fission ◽  
Pebble Bed ◽  
Fission Power ◽  
Triso Particles ◽  
Fuel Particles ◽  
Sem Investigation ◽  
High Temperature Gas ◽  
Test Fuel

There is currently renewed interest in high temperature nuclear fission power reactors. The Pebble Bed Modular Reactor (PBMR) is one of several high temperature gas-cooled reactors being investigated by researchers. The South African design of the PBMR is based on the original German design, with the fuel particles (called TRISO particles) being small multilayer spheres.

Burnup characteristic of CERMET and TRISO fuel particles used in Boiling Water Reactor with Pebble-Bed Coated Particles

2014 ◽  
Vol 2014 (1) ◽  
pp. 17-22
Author(s):  
Abdelfettah Benchrif ◽  
◽  
Abdelouahed Chetaine ◽  
Hamid Amsil ◽  
◽  
...  
Keyword(s):  
Boiling Water ◽  
Boiling Water Reactor ◽  
Pebble Bed ◽  
Water Reactor ◽  
Coated Particles ◽  
Triso Fuel ◽  
Fuel Particles

Measurement of kernel swelling and buffer densification in irradiated UCO and UO2 TRISO fuel particles from AGR-2

2019 ◽  
Author(s):  
John D Stempien ◽  
Mitchell A Plummer ◽  
Jason L Schulthess ◽  
Paul A Demkowicz
Keyword(s):  
Triso Fuel ◽  
Fuel Particles

scholarly journals Uniformity Assessment of TRISO Fuel Particle Distribution in Spherical HTGR Fuel Element Using Voronoi Tessellation and Delaunay Triangulation

2018 ◽  
Vol 2018 ◽  
pp. 1-6 ◽  
Author(s):  
Libing Zhu ◽  
Xincheng Xiang ◽  
Yi Du ◽  
Gongyi Yu ◽  
Ziqiang Li ◽  
...  
Keyword(s):  
Fuel Element ◽  
Delaunay Triangulation ◽  
Nearest Neighbor ◽  
Voronoi Tessellation ◽  
Fission Products ◽  
Fuel Particle ◽  
Volume Distribution ◽  
Triso Fuel ◽  
Triso Particles ◽  
Log Normal

Nonuniform distribution of tri-structural-isotropic (TRISO) fuel particles in a spherical fuel element (SFE) may increase the failure probability of the SFE in the high-temperature gas-cooled reactor, leading to the release of fission products. To evaluate the uniformity of the TRISO particles nondestructively, 3-dimensional cone-beam computed tomography is used to image the SFE, and TRISO particles are segmented. After TRISO particle positions are identified, the Voronoi tessellation and Delaunay triangulation are used to extract several geometric metrics. Results indicate that both the Voronoi volume distribution and the nearest neighbor-distance distribution follow the log-normal distributions, which provide strong evidence that the TRISO particles are approximately randomly uniformly distributed. Further study will be focused on validating the conclusion with more SFE data.

scholarly journals Oxidation Behavior of Matrix Graphite and Its Effect on Compressive Strength

2017 ◽  
Vol 2017 ◽  
pp. 1-6 ◽  
Author(s):  
Xiangwen Zhou ◽  
Cristian I. Contescu ◽  
Xi Zhao ◽  
Zhenming Lu ◽  
Jie Zhang ◽  
...  
Keyword(s):  
Compressive Strength ◽  
Pore Formation ◽  
Oxidation Behavior ◽  
Mercury Porosimetry ◽  
Strength Loss ◽  
Nuclear Graphite ◽  
Matrix Graphite ◽  
Oxidation Performance ◽  
Lower Oxidation ◽  
High Temperature Gas

Matrix graphite (MG) with incompletely graphitized binder used in high-temperature gas-cooled reactors (HTGRs) is commonly suspected to exhibit lower oxidation resistance in air. In order to reveal the oxidation performance, the oxidation behavior of newly developed A3-3 MG at the temperature range from 500 to 950°C in air was studied and the effect of oxidation on the compressive strength of oxidized MG specimens was characterized. Results show that temperature has a significant influence on the oxidation behavior of MG. The transition temperature between Regimes I and II is ~700°C and the activation energy (Ea) in Regime I is around 185 kJ/mol, a little lower than that of nuclear graphite, which indicates MG is more vulnerable to oxidation. Oxidation at 550°C causes more damage to compressive strength of MG than oxidation at 900°C. Comparing with the strength of pristine MG specimens, the rate of compressive strength loss is 77.3% after oxidation at 550°C and only 12.5% for oxidation at 900°C. Microstructure images of SEM and porosity measurement by Mercury Porosimetry indicate that the significant compressive strength loss of MG oxidized at 550°C may be attributed to both the uniform pore formation throughout the bulk and the preferential oxidation of the binder.

Fission Product Release Behavior of Individual Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

2000 ◽  
Vol 131 (1) ◽  
pp. 36-47 ◽  
Author(s):  
Kazuo Minato ◽  
Kazuhiro Sawa ◽  
Toshio Koya ◽  
Takeshi Tomita ◽  
Akiyoshi Ishikawa ◽  
...  
Keyword(s):  
High Temperature ◽  
Fission Product ◽  
Release Behavior ◽  
Product Release ◽  
Fuel Particles ◽  
High Temperature Gas

Methodology Development for Transient Flow Distribution Analysis in High Temperature Gas-Cooled Reactor

2020 ◽  
Author(s):  
Takeshi Aoki ◽  
Hiroyuki Sato ◽  
Hirofumi Ohashi
Keyword(s):  
High Temperature ◽  
Hydraulic System ◽  
System Analysis ◽  
Molecular Diffusion ◽  
Transient Flow ◽  
Flow Distribution ◽  
Distribution Analysis ◽  
Design Improvement ◽  
Air Ingress ◽  
High Temperature Gas

Abstract In the thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR), unintended flows such as gap flows between columns, cross flows between column layers and gap flows between permanent reflectors should be analyzed to minimizing the unintended flows. The flow distribution considering unintended flows in the reactor has been evaluated for steady and conservative condition. On the other hand, the transient thermal hydraulic analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to improve the thermal hydraulic system analysis code developed by Japan Atomic Energy Agency based on the RELAP5/MOD3 code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents for its design improvement utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated to evaluate the detailed flowrate distribution considering the unintended flows in the core and the molecular diffusion that is important to analyze beginning air ingress behavior in an air ingress accident triggered by a rupture of a primary coolant piping in HTGR. It is concluded that a prospect has confirmed to apply the improved thermal hydraulic system analysis code for transient flow distribution analysis for prismatic-type HTGRs.

scholarly journals Investigation on neutronic properties of ZrC coated advanced TRISO fuel for high-temperature gas-cooled reactors

2020 ◽  
Vol 1436 ◽  
pp. 012036
Author(s):  
F Aziz ◽  
M Panitra ◽  
A K Rivai ◽  
M Silalahi ◽  
N Sabrina ◽  
...  
Keyword(s):  
High Temperature ◽  
Triso Fuel ◽  
High Temperature Gas

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.

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