Papers by Keyword: Spacer Grid

Abstract: The spacer grids are part of the Fuel Element (FE) set of the PWR (Pressurized Water Reactor) type reactor. These grids maintain the position of the fuel rods within the arrangement of the FE, conserving among them the spacing necessary for the operation of the reactor. The grids are manufactured from the union of the intersecting points of stamped strips of Base Material (BM) Inconel 718, by a joint process called brazing. The addition metal (AM) used consists of a brazing paste based on Ni-Cr-P (nickel-chromium-phosphorus), which is added dropwise in the intersection of grids with a clearance of 0.025 mm. For this purpose, Inconel 718 smooth strips of 0.35 mm thick nickel plated samples were prepared, the AM was added and the vacuum oven was 10^-3 mbar, in different time and temperatures. The samples were prepared by metallography and characterized using optical microscopy (OM), scanning electron microscopy (SEM) and mechanical microhardness test. The purpose of this work is to characterize the brazing joint by means evaluate the size of precipitates and the different compounds formed in the joint. It was found different precipitates composed mainly of titanium and niobium. Phosphorus rich phases were found in the brazed region. The mean hardness of the BM was 469 ± 12 HV and in the joint region hardness of 1345 ± 34 HV was found in the lighter phases.

Abstract: Recently, a dual-cooled fuel (i.e. annular fuel) which is compatible with current operating PWR plants has been proposed in order to increase both power densities and safety margins. Due to the design concept that is compatible with current PWR plants, however, when compared with a current solid nuclear fuel it shows a narrow gap between fuel rods and needs to modify spacer grid shapes and their positions. Because a flow-induced vibration by fast primary coolant is inevitable phenomenon, it is necessary to examine the fretting wear behavior between an annular fuel and designed spacer grids. In this study, fretting wear has been performed to evaluate the wear resistance of the annular fuel by using specially designed spring and dimple of spacer grids that have a cantilever type and a hemispherical shape, respectively. At the spring specimen with relatively small stiffness value, fretting wear was initiated at both end regions and then proceeded gradually to center region. Based on the test results, the fretting wear behavior of annular fuel was compared with the current solid nuclear fuel and a comparative factor of its reliability was proposed.

Abstract: Safety increasing of nuclear reactors is urgent problem for atomic industry. It takes 100% noncontact dimensional inspection of nuclear reactor components. Optoelectronic methods and systems for 3D inspection of fuel assembly elements, including spacer grids and fuel elements, are presented. The universal structured light method for 3D inspection of Russian and western grid spacers using diffraction elements is developed. For fuel elements inspection one considers the shadow method which allows one to implement full inspection of articles under transported along the automatic production line. TDI SIE has developed the commercial version of systems for integral inspection of spacer grids and fuel elements. The experimental results of industrial testing are given.

Abstract: This study contains several estimation methods of the static buckling load for the spacer grid of nuclear fuel assembly in pressurized water reactor. Three different estimation methods were proposed for the calculation of the static buckling loads of spacer grid. The linear and non-linear static buckling analyses were performed to estimate the static buckling load of the spacer grids using ANSYS program. The analyses results were compared with the static buckling test results. Based on the analysis and test results, the applicability of the proposed estimation method for the static buckling load of the spacer grid was investigated.

Abstract: An effort has been made in the present investigation to evaluate the wear resistance of nuclear fuel rods with a variation of the supporting spring shapes and their stiffness by conducting fretting wear tests in room temperature air and water. With increasing slip amplitude, the wear volume and maximum wear depth are increased with increasing slip amplitude. However, these are not linearly increased with increasing spring stiffness. After the wear test, the worn surfaces were observed to investigate the debris behavior and wear mechanism by using an optical microscope (OM). The results indicated that almost all of the wear debris remained between the contacting surfaces and the wear debris layers were well developed in room temperature air. Besides, some of the debris also remained on the worn surface in room temperature water. This result shows that the remaining debris effect on the worn surface was more dominant than the spring stiffness one. So, in order to improve the fretting wear resistance of a nuclear fuel rod, it is necessary to consider the debris behavior between contacting surfaces even though the supporting spring shape was optimized by considering the contact mechanics, material compatibility, etc. From the experimental results, the fretting wear mechanisms and the effect of spring properties were discussed.

Abstract: In this study, the variation of spring characteristics with increasing temperature was examined and the effect of their variations on the wear behavior of a nuclear fuel rod in both room and high temperature (300°C) water conditions was evaluated. From the results of the load-displacement tests, the spring stiffness was remarkably varied with increasing temperature. The results of the wear tests indicated that the wear damages are decreased at high temperature water when compared with the room temperature result. These results indicated that the removal mechanisms of wear debris at high temperature water are dependent on not only the formation of the wear particle layer but also on the changed contact conditions such as the contact length or area due to the stiffness drops.