Formation of Pellet-Cladding Bonding Layer in High Burnup BWR Fuels (original) (raw)
Related papers
On the origins and the evolution of the fuel-cladding bonding phenomenon in PWR fuel rods
Journal of Nuclear Materials, 2019
This study proposes a new and detailed description of the fuel-cladding bonding phenomenon occurring in PWR fuel rods. Very early and late bonding states were characterized on specimens of 35.3 GWd.t U À1 moderate burnup and of 64.5 GWd.t U À1 high burnup respectively. Results were then compared with those achieved on a recreated bonded situation obtained on a Zircaloy-4/hyper-stoichiometric UO 2þx model materials diffusion couple. These results tend to indicate that a chemical adhesion is probably at the origin of the PWR fuel-cladding bonding. In addition, the progressive formation of ZrO 2 /UO 2 interfacial circumvolutions observed with increasing burnup, which lead to the physical anchorage of ZrO 2 and UO 2 , is likely to lead to their mechanical adhesion. Thus, the in-reactor ZrO 2 /UO 2 bonding could be considered, since the occurrence of the first bonded situations, as an adhesion phenomenon owning two components: an initial chemical progressively strengthened by a second mechanical.
Fuel oxidation and irradiation behaviors of defective BWR fuel rods
Journal of Nuclear Materials, 1995
The fuel oxidation of UO 2 pellets in two types of defective Zircaloy-clad BWR fuel rods with small leaks has been examined along both pellet diametral and fuel rod axial directions. The post-defect irradiation time was a few months for the base-irradiated full length rod and several minutes for the power-ramped segment rod. No phase change to higher order oxides of U409 or U30 8 was found, but hyperstoichiometric UO2+ x with fluorite structure was still present for both fuels. The fuel oxidation significantly depended on the defect size and distance from the defect. The pellet volume-averaged O/M ratios at various axial locations were in the range of 2.02-2.06 for the base-irradiated fuel, and about 2.01 for the power-ramped fuel. The data revealed that pellet oxidation by steam proceeded notably even in a short period of several minutes and played a more important role for generating liberated hydrogen, which could cause secondary hydriding of Zircaloy cladding, in comparison with the inner wall oxidation of cladding.
Fuel R&D at Studsvik II. General studies of fuel behaviour including pellet-cladding interaction
Nuclear Engineering and Design, 1997
An example of general fuel behaviour studies on power reactor fuel is described. Examples of recent non-destructive and destructive hot cell investigation techniques are given, illustrated by an investigation of full-size power reactor fuel rods and by a detailed study of fission product redistribution and release. A review is given of some international research and development projects addressing the pellet-cladding interaction problem. © 1997 Elsevier Science S.A.
Proceedings of 18th …, 2005
Several high-burnup PWR-fuel-claddings associated to high corrosion levels have been subjected to RIA transients in the CABRI facility. The zirconia layer surrounding the Zircaloy-4 fuel rods exhibited a complex behavior during the RIA transients. Regularly spaced incipient cracks are initiated in the zirconia layer and stop at the Zircaloy interface beneath the oxide. These cracks are mainly initiated in the direction normal to the maximum principal stress. The crack density in post-test metallographies appears to be tightly linked to the maximum applied strain. This cracking process is sometimes followed by oxide spalling. Oxide spalling has a major impact on thermal-mechanical fuel rod behavior during such transients, thus the understanding and modeling of spallation is an issue that deserves attention. The present paper provides a quantitative analysis of the related data generated within the scope of the CABRI REP-Na tests, in the last decade, and more recently in the first tests of the CABRI Water loop program. A link with several other mechanical testing programs such as tensile tests, creep tests is also established.
Properties of the high burnup structure in nuclear light water reactor fuel
Radiochimica Acta
The formation of the high burnup structure (HBS) is possibly the most significant example of the restructuring processes affecting commercial nuclear fuel in-pile. The HBS forms at the relatively cold outer rim of the fuel pellet, where the local burnup is 2–3 times higher than the average pellet burnup, under the combined effects of irradiation and thermo-mechanical conditions determined by the power regime and the fuel rod configuration. The main features of the transformation are the subdivision of the original fuel grains into new sub-micron grains, the relocation of the fission gas into newly formed intergranular pores, and the absence of large concentrations of extended defects in the fuel matrix inside the subdivided grains. The characterization of the newly formed structure and its impact on thermo-physical or mechanical properties is a key requirement to ensure that high burnup fuel operates within the safety margins. This paper presents a synthesis of the main findings fro...
The assessment of the mechanical properties of the highly irradiated fuel claddings during an RIA (Reactivity Initiated Accident) has been carried out in the framework of the PROMETRA programme. Three main types of tests including burst tests, hoop and axial tensile tests, have been performed in CEA-Saclay hot laboratories in order to determine the cladding tensile properties used in the SCANAIR code. The representativeness of each test with regard to the RIA loading conditions can be addressed and analyzed in terms of strain or stress ratio. The present paper reports the high strain rate ductile mechanical properties of irradiated ZIRLO TM and M5 TM alloys derived from the PROMETRA program and their comparison to the stress-relieved irradiated Zircaloy-4. Results of specific analysis of the behaviour of the 6 cycle M5 TM and ZIRLO TM 75 GWd/tM for temperatures higher than 600°C are also presented.
Behavior of unirradiated Zr based uranium metal fuel under reactivity initiated accident conditions
Nuclear Engineering and Design, 2008
Reactivity initiated accident (RIA) tests with 4 unirradiated Zr based uranium metal fuel rods were performed to establish a criterion which should be observed under RIA conditions. Of the four tests, fuel failures were observed in the two tests that experienced the maximum energy depositions of 188 and 212 cal/g, respectively. However, the fuel failures were not observed at the place of a maximum energy deposition but at the position where the thermocouples were installed; one failed at the position whose local energy deposition was 150 cal/g, and the other one at the place with energy deposition of 170 cal/g. The fuel failures seem to have occurred because excessive pressure, which was caused by the partial melting of the fuel meat, was applied to the cladding with a reduced thickness. However, other parts of the fuel rods including the place of a maximum energy deposition maintained their integrity and a big change in the temperature and pressure in the internal capsule, which would be an indication of the fragmentation and dispersion of the fuel meat into the internal capsule, was not observed. Visual inspection also showed that, except for the thermocouple positions, there was no trace of clad failure such as the formation of brittle cracks in the cladding or melting of the cladding. Therefore, for the Zr based uranium metal fuel rods, it can be concluded that the threshold energy deposition above which fragmentation and dispersion of fuel meat into the primary coolant system is expected to occur could be higher than 212 cal/g.
Telematika, 2011
In recent years, a novel class of zirconium alloys having the melting temperature of 990-1160 K has been developed. Based on novel zirconium matrix alloys, high uranium content fuel pin with U-9Mo has been developed according to capillaryimpregnation technique. The pin shows its thermal conductivityranging from 18 to 22 w/m/K that is comparably higher than UO2pellet pin. The paper presents the met-met fabrication and thermal performance analysis of the fuel in typical PWR. Thefabrication consists of mixing UO2 powder or granules and a novel Zr-alloy powder having low melting point, filling the mixture in a cladding tube that one of its end has been plugged, heating the pin to above melting temperature of Zr-alloy for an hour, natural cooling and heat treating at 300 K for ½ hr. The thermal analysis takes into account the pore and temperature distribution and high burn up effect to pellet conductivity. The thermal diffusivity ratio of novel to conventional fuel has been used as correction factor for the novel fuel conductivity. The results show a significant lowering pellet temperature along the radius until 1000 K at the hottest position. The analysis underestimates since the gap conductivity has been treated as decreased by 2 % fission gas released that is not real since the use of lower temperature, and also decreasing thermal conductivity by porosity formation will much lower. The analysis shows that the novel fuel has very good thermal properties which able to pass the barrier of 65 MWD/kg-U, the limit to day commercial fuel. The burn-up extension means less fresh fuel is needed to produce electricity, preserve natural uranium resource, easier fuel handling operational per energy produced.
An Investigation of Liquefaction in Irradiated TRIGA Fuel Exposed to Relatively High Temperatures
2020
This report presents the findings of an investigation into high-temperature fuel cladding chemical interactions (FCCI) in Training, Research, Isotopes, General Atomics (TRIGA) fuel rods. A TRIGA fuel-rod core or meat is principally composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. The fuel is clad in sealed 304SS or Incoloy 800 tubes. At high temperatures, the fuel will interact with the cladding, resulting in FCCI.