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].

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.

Investigation on Iodine Release Behavior During the Operation of High Temperature Engineering Test Reactor (HTTR)

2013 ◽  
Author(s):  
Shohei Ueta ◽  
Hiroyuki Inoi ◽  
Yoshitaka Mizutani ◽  
Hirofumi Ohashi ◽  
Jin Iwatsuki ◽  
...  
Keyword(s):  
High Temperature ◽  
Local Area ◽  
Release Behavior ◽  
Primary Coolant ◽  
Energy Agency ◽  
Xenon Isotopes ◽  
Test Reactor ◽  
Forced Cooling ◽  
Iodine Release ◽  
Testing Plan

Japan Atomic Energy Agency (JAEA) has planned to investigate on iodine release behavior from fuel through the testing operation of High Temperature Engineering Test Reactor (HTTR) in order to contribute to the reasonable estimation of the radiation exposure necessary for the realization of HTGR in the future. In this test, the fractional release of iodine will be measured and evaluated by measuring xenon isotopes, the daughter nuclides of iodine isotopes, in the primary coolant sampling under the loss-of-forced cooling (LOFC) test by which the primary coolant circulator is shut down and/or the manual scram test of HTTR. In parallel, the local area of primary coolant circuit where iodine is plated-out will be evaluated. This paper describes the testing plan and the preliminary analytical study on the release behavior of iodine and xenon isotopes through the operation of HTTR.

scholarly journals Interaction of Noble Metal Fission Products with Pyrolytic Silicon Carbide

1982 ◽  
Vol 40 ◽  
pp. 566-567
Author(s):  
R. J. Lauf ◽  
D. N. Braski
Keyword(s):  
Silicon Carbide ◽  
Fission Product ◽  
Noble Metals ◽  
Fission Products ◽  
Pyrolytic Carbon ◽  
Enriched Uranium ◽  
Fuel Particles ◽  
The Many ◽  
Sic Coatings ◽  
Palladium Group

Fuel particles for the High-Temperature Gas-Cooled Reactor (HTGR) contain layers of pyrolytic carbon and silicon carbide, which act as a miniature pressure vessel and form the primary fission product barrier. Of the many fission products formed during irradiation, the noble metals are of particular interest because they interact significantly with the SiC layer and their concentrations are somewhat higher in the low-enriched uranium fuels currently under consideration. To study fission product-SiC interactions, particles of UO2 or UC2 are doped with fission product elements before coating and are then held in a thermal gradient up to several thousand hours. Examination of the SiC coatings by TEM-AEM after annealing shows that silver behaves differently from the palladium group.

Diffusion of Silver in Single Crystalline 6H-SiC

2008 ◽  
Author(s):  
T. T. Hlatshwayo ◽  
N. G. van der Berg ◽  
E. Friedland ◽  
J. B. Malherbe ◽  
P. Chakraborty
Keyword(s):  
Silicon Carbide ◽  
High Temperature ◽  
Nuclear Reactor ◽  
Fission Products ◽  
Pyrolytic Carbon ◽  
Reactor Safety ◽  
Fission Fragments ◽  
Coated Particle ◽  
Nuclear Reactor Safety ◽  
Fuel Elements

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].

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