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Thermo-chemical-mechanical modeling of nuclear fuel behavior : Impact of oxygen transport in the fuel on Pellet Cladding Interaction

Abstract : The goal of this thesis is to study the impact of oxygen transport on thermochemistry of nuclear fuel and pellet cladding interaction. During power ramps, nuclear fuel is exposed to high temperature gradients. It undergoes chemical and structural changes. The fuel swelling leads to a mechanical contact with the cladding causing high mechanical stresses in the cladding. Simultaneously, chemically reactive gas species are released from the hot pellet center and can interact with the cladding. The combination of these chemical and mechanical factors may lead to the cladding failure by iodine stress corrosion cracking. It has been proven that oxygen transport under high temperature gradients affects irradiated fuel thermochemistry, a phenomenon which may be of importance for stress corrosion cracking. This thesis presents 3D simulations of power ramps in pressurized water reactors with the fuel performance code ALCYONE, which is part of the computing environment PLEIADES. The code has been upgraded to couple the description of irradiated fuel thermochemistry already available with oxygen transport taking into account oxygen thermal diffusion. The impact of oxygen redistribution during a power transient on irradiated fuel thermochemistry in the fuel and on chemically reactive gas release from the fuel (consisting of I(g), I2(g), CsI(g), TeI2(g), Cs(g) and Cs2(g), mainly) is studied. The simulations show that oxygen redistribution, even if moderate in magnitude, leads to the reduction of metallic oxides (molybdenum dioxide, cesium molybdate, chromium oxide) at the fuel pellet center and consequently to the release of a much greater quantity of gaseous cesium, in agreement with post-irradiation examinations. The three-dimensional calculations of the quantities of importance for iodine stress corrosion cracking (hoop stress, hoop strain, iodine partial pressure at the clad inner wall) are then used in simulations of clad crack propagation.
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Submitted on : Tuesday, May 12, 2020 - 9:08:09 AM
Last modification on : Tuesday, October 20, 2020 - 11:03:22 AM


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  • HAL Id : tel-02570386, version 1


Piotr Konarski. Thermo-chemical-mechanical modeling of nuclear fuel behavior : Impact of oxygen transport in the fuel on Pellet Cladding Interaction. Materials. Université de Lyon, 2019. English. ⟨NNT : 2019LYSEI080⟩. ⟨tel-02570386⟩



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