Development and benchmarking of a 2D transient thermal model for GFR plate-type fuel

Rakesh Chawla, Konstantin Mikityuk, Petr Petkevich
Elsevier Science Ltd., 2007
Journal paper

Abstract

Several new fuel designs are currently being considered within the generation-IV gas-cooled fast reactor (GFR) project. One of them, which is the GFR reference design, is a ceramic plate matrix with a honeycomb inner structure containing small fuel cylinders. The fuel is mixed plutonium-uranium carbide, while the matrix material is silicon carbide. A two-dimensional approach to simulation of the thermal behaviour (for both steady-state and transient analysis) of such fuel, taking into account the inner heterogeneity, has been developed and benchmarked against detailed finite-element calculations. The resultant model can provide reliable fuel and matrix temperatures to determine neutronic feedbacks, as also to evaluate fuel integrity. As such, it has been integrated into PSIs coupled code system FAST, which aims at the comprehensive safety analysis of advanced fast reactor systems.[All rights reserved Elsevier].

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The reference fuel design currently being considered within the Generation-IV Gas-cooled Fast Reactor (GFR) project is a ceramic plate matrix with a honeycomb inner structure containing small fuel cylinders. The fuel is mixed plutonium-uranium carbide, while the matrix material is silicon carbide. The present paper describes the mechanical part of a thermal-mechanical model being developed for studying the transient behavior of this highly heterogeneous fuel type. Benchmarking has been carried out against detailed finite-elements modeling (FEM). The resultant thermal-mechanical model can provide reliable fuel and cladding (matrix) stress/strain conditions to evaluate temperatures and neutronic feedbacks. As such, it has been integrated into PSI's coupled code system FAST, which aims at the comprehensive safety analysis of advanced fast reactor systems. The detailed FEM analysis of the GFR fuel has been useful not only for benchmarking the new model, but also for obtaining an in-depth understanding of fuel stress/strain characteristics, which cannot be reproduced with simplified models. Thereby, the range of applicability of the new model has clearly been defined. In particular, the 3D FEM analysis has revealed a concentration of stresses at the pellet corners during pellet/matrix contact, which could lead to fuel element failure. This effect is found to be mitigated considerably, if the fuel pellets are shaped in a manner which enhances the contact area. [All rights reserved Elsevier].

Elsevier Science Ltd.2009

Calculations of Reactivity-Initiated Transients in Gas-Cooled Fast Reactors Using the Code System FAST

Konstantin Mikityuk

The FAST code system is a general tool for analyzing advanced reactors from the viewpoint of the static and dynamic behavior of the whole reactor system. It includes an integrated three-dimensional representation of the core neutronics, appropriate modeling of the core thermal-hydraulics and fuel pin behavior, coupled to models of the reactor primary and secondary systems. Use is made largely of well-established individual neutronic, thermal-hydraulic and fuel behavior modules. Clearly, it is important to verify the individual parts of the code, including the links between them. The paper is focused on this detailed verification procedure. Steady-state conditions, as well as the transient behavior of hypothetical reactivity-initiated accidents, are investigated for two specific gas-cooled fast reactors. While the first system, a CO2-cooled CAPRA-CADRA core, is loaded with Superphénix-like MOX fuel, the second system being analyzed, a He-cooled Generation IV-like core, uses ceramic (U,Pu)C fuel dispersed in a silicon-carbide matrix. In the current study, the TRAC/PARCS elements of FAST are compared with the 3D-kinetics stand-alone ERANOS/KIN-3D code, which is considered state-of-the-art, using as far as possible equivalent options. A new methodology is proposed to improve a diffusion-theory, coarse-group PARCS-solution by scaling the original cross-section derivatives and input kinetic parameters.

Elsevier2006