Preliminary three-dimensional neutronic analysis of IFBA coated TRISO fuel particles in prismatic-core advanced high temperature reactor

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].

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Computational Model for the Neutronic Simulation of Pebble Bed Reactor’s Core Using MCNPX

International Journal of Nuclear Energy ◽

10.1155/2014/279073 ◽

2014 ◽

Vol 2014 ◽

pp. 1-12 ◽

Cited By ~ 3

Author(s):

J. Rosales ◽

A. Muñoz ◽

C. García ◽

L. García ◽

C. Brayner ◽

...

Keyword(s):

High Temperature ◽

Computational Model ◽

Inherent Safety ◽

Pebble Bed ◽

Pebble Bed Reactor ◽

Triso Fuel ◽

Very High Temperature Reactor ◽

Fuel Design ◽

High Temperature Reactor ◽

Very High

Very high temperature reactor (VHTR) designs offer promising performance characteristics; they can provide sustainable energy, improved proliferation resistance, inherent safety, and high temperature heat supply. These designs also promise operation to high burnup and large margins to fuel failure with excellent fission product retention via the TRISO fuel design. The pebble bed reactor (PBR) is a design of gas cooled high temperature reactor, candidate for Generation IV of Nuclear Energy Systems. This paper describes the features of a detailed geometric computational model for PBR whole core analysis using the MCNPX code. The validation of the model was carried out using the HTR-10 benchmark. Results were compared with experimental data and calculations of other authors. In addition, sensitivity analysis of several parameters that could have influenced the results and the accuracy of model was made.

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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

Advances in Science and Technology ◽

10.4028/www.scientific.net/ast.45.1944 ◽

2006 ◽

Vol 45 ◽

pp. 1944-1951 ◽

Cited By ~ 2

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.

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