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Experimental simulation of activity release from leaking fuel rods
Barbara Somfai,Zoltan Hozer,Imre Nagy 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.7
The Leaking Fuel Experiment test facility was designed to simulate the activity release from spent leakingfuel rods under steady state and transient conditions in the spent fuel pool. The experimental rigincluded an electrically heated fuel rod with different defects and a cooling system. The fission producttransport was simulated by potassium-chloride. The conductivity changes of the water in the coolingsystem were measured to provide information about the amount of released solution. Defects of differentsizes and positions were applied, together with a wide range of rod powers to simulate decay heat. Theproduced data can be used for predicting the activity release from leaking fuel under storage conditionsand for the interpretation of fuel examination procedures
M. Kir aly,Z. Hozer,M. Horvath,T. Novotny,E. Perez-Fero,N. Ver 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.2
The mechanical and corrosion behavior of the Russian zirconium fuel cladding alloy E110, predominantlyused in VVERs, has been investigated for many decades. The recent commercialization of a new, optimizedE110 alloy, produced on a sponge zirconium basis, gave the opportunity to compare the mechanicalproperties of the old and the new E110 fuel claddings. Axial and tangential tensile test experiments were performed with samples from both claddings in theMTA EK. Due to the anisotropy of the cladding tubes, the axial tensile strength was 10e15% higher thanthe tangential (measured by ring tensile tests). The tensile strength of the new E110G alloy was 11%higher than that of the E110 cladding at room temperature. Some samples underwent chemical treatment e slight oxidation in steam or hydrogenation e or heattreatment e in argon atmosphere at temperatures between 600 and 1000 C. The heat treatment duringthe oxidation had more significant effect on the tensile strength of the claddings than the oxidation itself,which lowered the tensile strength as the thickness of the metal decreased. The hydrogenation of thecladding samples slightly lowered the tensile strength and the samples but they remained ductile even atroom temperature.
Evaluation of axial and tangential ultimate tensile strength of zirconium cladding tubes
Marton Kiraly,Daniel Mihaly Antok,Laszlone Horvath,Zoltan Hozer 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.3
Different methods of axial and tangential testing and various sample geometries were investigated, andnew test geometries were designed to determine the ultimate tensile strength of zirconium claddingtubes. The finite element method was used to model the tensile tests, and the results of the simulationswere evaluated. Axial and tangential tensile tests were performed on as-received and machined fuelcladding tube samples of both E110 and E110G Russian zirconium alloys at room temperature to comparetheir ultimate tensile strengths and the different sample preparation methods
Segmented mandrel tests of as-received and hydrogenated WWER fuel cladding tubes
Kiraly, Marton,Horvath, Marta,Nagy, Richard,Ver, Nora,Hozer, Zoltan Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.9
The mechanical interaction between the fuel pellet and the cladding tube of a nuclear fuel rod is a very important for safety studies as this phenomenon could lead to fuel failure and release of radioactivity. To investigate the ductility of cladding tubes used in WWER type nuclear power plants, several mandrel tests were performed in the Centre for Energy Research (EK). This modified mandrel test was used to model the mechanical interaction between the fuel pellet and the cladding using a segmented tool. The tests were conducted at room temperature and at 300 ℃ with inactive as-received and hydrogenated cladding ring samples. The results show a gradual decrease in ductility as the hydrogen content increases, the ductile-brittle transition was seen above 1500 ppm hydrogen absorbed.