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

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

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.

scholarly journals Local atomic structure of Pd and Ag in the SiC containment layer of TRISO fuel particles fissioned to 20% burn-up

2018 ◽  
Vol 500 ◽  
pp. 316-326 ◽  
Author(s):  
Rachel L. Seibert ◽  
Kurt A. Terrani ◽  
Daniel Velázquez ◽  
John D. Hunn ◽  
Charles A. Baldwin ◽  
...  
Keyword(s):  
Atomic Structure ◽  
Local Atomic Structure ◽  
Triso Fuel ◽  
Fuel Particles

Development of High Temperature Gas-Cooled Reactor Fuel for Extended Burnup

2010 ◽  
Author(s):  
Shohei Ueta ◽  
Jun Aihara ◽  
Masaki Honda ◽  
Noboru Furihata ◽  
Kazuhiro Sawa
Keyword(s):  
High Temperature ◽  
Coating Layer ◽  
Fission Products ◽  
Pyrolytic Carbon ◽  
Principal Function ◽  
Energy Agency ◽  
Triso Fuel ◽  
Coating Layers ◽  
Test Reactor ◽  
Fuel Particles

Current HTGRs such as the High Temperature Engineering Test Reactor (HTTR) of Japan Atomic Energy Agency (JAEA) use Tri-Isotropic (TRISO)-coated fuel particles with diameter of around 1 mm. TRISO fuel consists of a micro spherical kernel of oxide or oxycarbide fuel and coating layers of porous pyrolytic carbon (buffer), inner dense pyrolytic carbon (IPyC), silicon carbide (SiC) and outer dense pyrolytic carbon (OPyC). The principal function of these coating layers is to retain fission products within the particle. Particularly, the SiC coating layer acts as a barrier against the diffusive release of metallic fission products and provides mechanical strength for the particle [1].

scholarly journals Long-Term Thermal Testing of Irradiated TRISO Fuel Particles from the AGR-2 Experiment

2019 ◽  
Author(s):  
Z. Burns ◽  
T. Gerczak ◽  
J. Hunn ◽  
D. Skitt
Keyword(s):  
Thermal Testing ◽  
Triso Fuel ◽  
Fuel Particles

Preliminary Development of a TRISO Fuel Performance Analysis Code: FFAT

2018 ◽  
Author(s):  
Jian Li ◽  
Ding She ◽  
Lei Shi ◽  
Jing Zhao
Keyword(s):  
Performance Analysis ◽  
Mechanical Performance ◽  
Performance Model ◽  
Fuel Performance ◽  
Triso Fuel ◽  
Coated Particle ◽  
First Results ◽  
Fuel Particles ◽  
Accident Conditions ◽  
Radioactivity Retention

Tristructural isotropic (TRISO) fuel particles are chosen as the major fuel type of High temperature gas cooled reactor (HTGR). The TRISO coated particle also acts as the first barrier for radioactivity retention. The performance of the TRISO coated particle has a significant influence on the safety of HTGR. A set of fuel performance analysis codes have been developed during the past decades. The main functions of these codes are conducting stress calculation and failure probability prediction. PANAMA is a widely used German version fuel performance analysis code, which simulates the mechanical performance of TRISO coated particle under normal and accident conditions. In this code, only a simple pressure vessel model is considered, which is insufficient in stress analysis and fuel failure rate prediction. Nowadays, efforts have been done to update the fuel performance model utilized in PANAMA code, and a new TRISO fuel performance analysis code, FFAT, is under developed. This paper describes the newly updated TRISO fuel performance model and presents some first results based on the updated model.

Thermodynamic Modelling of the Uranium-Carbon-Oxygen System in the Frame of the Uranium Oxide and Carbon Interaction in the TRISO Fuel Particle of High Temperature Reactor

2006 ◽  
Vol 45 ◽  
pp. 1944-1951 ◽  
Author(s):  
Jean Christophe Dumas ◽  
Jean Paul Piron ◽  
Sylvie Chatain ◽  
Christine Guéneau
Keyword(s):  
High Temperature ◽  
Fission Products ◽  
Nuclear Materials ◽  
Chemical Compositions ◽  
Fuel Particle ◽  
Chemical Systems ◽  
Triso Fuel ◽  
Physico Chemical ◽  
Chemical Studies ◽  
Fuel Particles

A thermodynamic approach is necessary in order to predict and understand physico-chemical phenomena occurring in nuclear materials under irradiation, involving large chemical systems with a lot of elements including both initial nuclides and fission products (FP). In the frame of thermo-chemical studies of the High Temperature Reactors fuel, a first step is to assess the (U-O-C) system in order to understand the interaction between the UO2 kernel and the pyrocarbon layers constituting such a fuel particle. Our model for irradiated oxide fuel, based on Lindemer’s analysis, has been improved by introducing the (U-O-C) model developed by C. Guéneau & al into the SAGE code. Chemical compositions and related carbon oxides pressures of irradiated TRISO fuel particles have been calculated with the data published by Minato & al. We discuss our results by comparison with their thermochemical calculations and with their experimental observations. This approach can be used to predict the behaviour of complex nuclear materials, especially for the different kind of fuel materials considered in the frame of Gas Fast Reactors.

scholarly journals Modeling Strategies to Compute Natural Circulation Using CFD in a VHTR After a LOFA

2012 ◽  
Author(s):  
Yu-Hsin Tung ◽  
Richard W. Johnson ◽  
Ching-Chang Chieng ◽  
Yuh-Ming Ferng
Keyword(s):  
Heat Transfer ◽  
Natural Circulation ◽  
Department Of Energy ◽  
Full Potential ◽  
The Core ◽  
Triso Fuel ◽  
Flow And Heat Transfer ◽  
Local Symmetries ◽  
Fuel Particles ◽  
Global And Local

A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted tristructural-isotropic (TRISO) fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coolant circulators are lost for some reason, causing a loss of forced coolant flow through the core. In such an event, it is desired to know what happens to the (reduced) heat still being generated in the core and if it represents a problem for the fuel compacts, the graphite core or the reactor vessel (RV) walls. One of the mechanisms for the transport of heat out of the core is by the natural circulation of the coolant, which is still present. It is desired to know how much heat may be transported by natural circulation through the core and upwards to the top of the upper plenum. It is beyond current capability for a computational fluid dynamics (CFD) analysis to perform a calculation on the whole RV with a sufficiently refined mesh to examine the full potential of natural circulation in the vessel. The present paper reports the investigation of several strategies to model the flow and heat transfer in the RV. It is found that it is necessary to employ representative geometries of the core to estimate the heat transfer. However, by taking advantage of global and local symmetries, a detailed estimate of the strength of the resulting natural circulation and the level of heat transfer to the top of the upper plenum is obtained.

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