[en] The restructuring process of the high burnup structure (HBS) formation in UO_2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO_2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO_2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results. - Highlights: • A model for evolution of dislocation density and grain size in HBS is proposed. • The dislocation can also be annealed when the temperature is high enough. • Original driving force for subdivision is mostly accumulation of dislocation loops. • The temperature threshold of the subdivision is predicted at 1300–1400 K.

[en] A model for the swelling behavior of fission gas in the range of low temperature of irradiated uranium dioxide fuel is established., In this paper, the finite difference method is adopted to compile a computational program and quantitatively calculated the fractions of fission gas as solid solution in UO₂, the density and average radius of intragranular bubbles and their contributions to the swelling of uranium dioxide fuel. in different bum-ups and temperatures of uranium dioxide fuel. The calculation shows that this model can be used to predict the fission gas swelling of uranium dioxide fuel versus burn-up in the range of low temperature. (authors)

[en] The adsorption sites and mechanism of H atom on α-Zr (0001) surface were calculated and analyzed from microstructure, adsorption probability, adsorption energy, Mulliken charge population and density of state, and etc. based on the integration of Monte Carlo (MC) simulation and first-principle density functional theory (DFT) method. The results indicated that the H atom firstly generated physical adsorption on the Zr (0001) surface and then changes to chemical adsorption. The charge continuously transferred from the surface Zr (0001) atoms to the H atom throughout the entire process, and finally stabilized. Furthermore, the H atom directly bonded with the most surface Zr (0001) atoms after stable adsorption, and the major contribution of Zr-H bond was made by partial density of state of H (s), Zr (s) and Zr (d) orbitals. Comprehensive analysis shows that the priority order of the adsorption sites of H atoms on the Zr (0001) surface is hexagonal close packed gap (hcp) > face centered cubic gap (fcc) > bridge, and the top site is the impossible adsorption site. (authors)

[en] An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated UO_2 fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO_2 fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO_2 fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data

[en] The correlative theories of the uranium dioxide fuels swelling driven by the precipitation of the insoluble fission products in the fuels matrix have been analyzed. After modified the Rest model, a suitable theoretical model has been established to carry out numerical simulation in the study of the irradiation-induced swelling in dispersion fuel. Based on the established Modified Rest model, a computational program has been compiled and then used to calculate the irradiation-induced swelling of the dispersion fuel especially the fission gas behavior in the same conditions quantitatively. The results indicate that the trends of the irradiation-induced swelling rate, density and radius of the bubbles as a function of burn-up and temperature are confirmed with the irradiation experimental data in comparison. When the burn-up is high enough, the irradiation induced defects led to the subdivision of the grains in fuel. The subgrain boundary net work promotes the migration of fission gas. This will result in large bubbles formed in subgrain boundary and then induced the accelerated rate of irradiation-induced swelling in fuel. This work paves the way for both the development of numerical computational program for the irradiation-induced swelling behavior in fuel and assessing physical property of the nuclear fuel elements. (authors)

[en] The Ag-In-Cd ternary alloy will transfer to Ag-In-Cd-Sn quaternary alloy under neutron irradiation in nuclear reactor due to the transmutation reactions inside the Ag-In-Cd ternary alloy. It is very important for recognizing the irradiation behavior of Ag-In-Cd alloy by research on the microstructures and thermo-physics properties of the Ag-In-Cd-Sn alloys with the different Sn content. The microstructures of Ag-In-Cd-Sn alloys with the different Sn content have been investigated by optical microscope (OM) and scanning electron microscope (SEM). The thermal diffusivity and conductivity have been measured by laser bombard method in this work. The results showed that the transformation induce chemical modifications inside the single phase alloy and, further, formation of two phases. The thermal diffusivity and conductivity of Ag-In-Cd-Sn alloy decreased as the Sn content increased under the uniform temperature but increased as the temperature increased under the uniform Sn content. (authors)

[en] Ni foil is used as interlayer to join Zircaloy-4 by transient liquid phase bonding. The Zone_β→α is formed in the joints and its width increases with temperature. The formation of Zone_β→α is controlled by both bonding temperature and Ni concentration. The Zone_β→α is mainly composed of primary Zr phase and eutectic Zr(Zr₂Ni) structure. The micro-hardness of the joints mainly depends on the precipitates and the size of α-Zr phase. The non-planar bonding interfaces are developed in the joints when the bonding temperature above 960 °C. Resulting from the non-planar bonding interfaces, the maximum shear strength of the joints is 358 ± 19 MPa

[en] Highlights: •A mechanistic model of fission gas swelling in LWR UO₂ fuel is proposed. •Fission fragments play the role of creator and destroyer of intragranular bubbles. •A modified equation of state for Xe is used instead of van der Waals equation. •The model correctly predicts the matrix swelling of UO₂ fuel up to 70 GW d/tU. -- Abstract: An understanding of the behavior of fission gas in uranium dioxide (UO₂) fuel is necessary for the prediction of the performance of fuel rods under irradiation. A mechanistic model for matrix swelling by the fission gas in LWR UO₂ fuel is presented. The model takes into account intragranular and intergranular fission gas bubbles behavior as a function of irradiation time, temperature, fission rate and burn-up. The intragranular bubbles are assumed to be nucleated along the track of fission fragments, which play the dual role of creator and destroyer of intragranular bubbles. The intergranular bubble nuclei is produced until such time that a gas atom is more likely to be captured by an existing nucleus than to meet another gas atom and form a new nucleus. The capability of this model was validated by a comparison with the measured data of fission gas behavior such as intragranular bubble size, bubble density and total fuel swelling. It was found that the calculated intragranular bubble size and density are in reasonable agreement with the measured results in a broad range of average fuel burn-ups 6–83 GW d/tU. Especially, the model correctly predicts the fuel swelling up to a burn-up of about 70 GW d/tU.

[en] Ag-In-Cd control rods are widely used in PWR nuclear power plants. The irradiation swelling behavior of Ag-In-Cd alloy is very important to the safety assessment of control rod during its operation, and the change of composition is the primary cause for the swelling. The relationship between the alloy composition and thermo neutron fluence was seldom reported up to today. In this work, a group of differential equations was proposed to describe the composition of Ag-In-Cd alloy during irradiation based on the transmutation reactions and the reaction cross sections. A numeric resolution to the equations was established. The alloy compositions at different thermo-neutron fluences were calculated and a group of formula for composition prediction was obtained. The average content for each element is approximately a linear function of thermo neutron fluence, and the content of Cd and Sn in the surface layer will be higher than two times of that in the center. (authors)

[en] AgInCd alloy is widely used as neutron absorber in nuclear reactors. However, the AgInCd control rods may fail during service due to the irradiation swelling. In the present study, a calculational method is proposed to calculate the composition change of the AgInCd absorber. Calculated results show that neutron fluence has significant impact on the chemical compositions. Ag and In contents gradually decrease while Cd and Sn conversely increases from the center to the rim of AgInCd absorber due to the depression of neutron flux. The composition change at the surface is higher almost two times than that at the center. Based on the calculated compositions, six simulated AgInCdSn alloys were prepared and examined. With the increase of Cd and Sn, the simulated AgInCdSn alloys transform from a single fcc phase into the mixed fcc and hcp phases, and finally into the single hcp phase. The atomic volume of the hcp phase is obviously larger than the fcc phase. The fcc-hcp transformation results in considerable volume swelling of the AgInCd absorber. Moreover, the lattice parameters of the fcc and hcp phases gradually increase with Cd and Sn contents, which also can induce small volume swelling