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.

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.

This publication sets out the findings of an international seminar, held in Aix-en-Provence, France in March 2004, which considered recent progress in the field of pellet-clad interaction in light water reactor fuels. It also reviews current understanding of relevant phenomena and their impact on the nuclear fuel rod under the widest possible conditions, and about both uranium-oxide and mixed-oxide fuels.

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr and U-Pu-Zr alloy fuel elements irradiated in the Experimental Breeder Reactor-II (EBR-II). The electrometallurgical treatment process extracts usable uranium from irradiated fuel elements and places residual fission products, actinides, process Zr, and cladding hulls (small segments of tubing) into two waste forms--a ceramic and a metal alloy. The metal waste form will contain the cladding hulls, Zr, and noble metal fission products, and it will be disposed of in a geologic repository. As a result, the expected composition of the waste form will need to be well understood. This report deals with the condition of the cladding, which will make up a large fraction of the metal waste form, after irradiation in EBR-II and before insertion into the electrorefiner. Specifically, it looks at layers that can be found on the inner surface of the cladding due to in-reactor interactions between the alloy fuel and the stainless steel cladding that occurs after the fuel has swelled and contacted the cladding. Many detailed examinations of fuel elements irradiated in EBR-II have been completed and are discussed in the context of interaction layer formation in irradiated cladding. The composition and thickness of the developed interaction layers are identified, along with the irradiation conditions, cladding type, and axial location on fuel elements where the thickest interaction layers can be expected to develop. It has been found that the largest interaction zones are observed at combined high power and high temperature regions of fuel elements and for fuel elements with U-Pu-Zr alloy fuel and D9 stainless steel cladding. The most prevalent, non-cladding constituent observed in the developed interaction layers are the lanthanide fission products.