Isıl-Nötronik Etkileşimlerin Yakıt Elemanlarının Tesir Kesitlerine Etkisi
Sarıcı Türkmen, Gülçin
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Accurate and precise neutronics calculations are important for reactor safety while defining safety limits for nuclear reactors. These calculations include the generation of neutron cross sections, as well. Some simplifications, assumptions and approaches are made during this process. Isothermal temperature approach is usually used in the fuel element since it reduces the computational cost. Such an approach leads to uncertainty on the reactor parameters. The effective cross-sections of the isotopes in the fuel vary directly with the temperature due to the Doppler effect, particularly in the resonance zone. Temperature-dependent cross section affects the self-shielding calculations. Therefore, use of the radial temperature profile in the fuel will give more accurate results. The aim of this thesis is to investigate the uncertainties due to the isothermal temperature approach in the fuel by solving heat equation with the space-dependent volumetric heat generation rate and temperature-dependent thermal conductivity coupled with the reactor physics code. The effect of the radial temperature profile on the reactor parameters was investigated by describing the fuel element with a multi-zone representation with different temperatures along the fuel radius. The DRAGON code is used in calculations. In the calculations, the effect of axial temperature change was not considered. The unit cells of the PWR TMI-1, BWR Peach Bottom and VVER-1000 Kozloduy reactors examined in the Uncertainty Analysis in Modeling (UAM) project were used as reference in the calculations. The results of the isothermal temperature approach were compared with the results of the temperature profile. The effects of temperature profile on criticality change, δk, Doppler coefficient, radial heat generation rates and temperature distributions were investigated. The results show that the magnitude of the change in k∞ depends directly on the reactor type, the average fuel temperature, the cross section of the library used, the solution method and other parameters used in the calculations. For three reactor types, there is an under estimation of approximately 110 pcm due to the use of the temperature profile and approximately 220 pcm due to the multi-region representation in k∞. Doppler coefficient of reactivity is overestimated to be 5-10%. As the average fuel temperature increases, the uncertainty in k∞ due to the use of the temperature profile also increases. An increase of 100 K in temperature is caused by a change of 15 pcm in k∞. The use of different neutron cross-section library leads to a change of maximum 5 pcm in k∞. There is no observed difference between the different versions of the same library. While examining effect of ENDF/B-VII.1 libraries with different energy group numbers, about 26-27 pcm in the k∞ value due to the use of the temperature profile and the multi-region representation is observed as the number of energy groups increases. In addition, the calculations made for the fuel assembly show that the full assembly can be deducted by modeling a fuel rod with average power, instead of modeling the full assembly. As a result of the calculations, the modeling of the fuel element with the temperature profile as the multi-region is required in order to accurately estimate the safety constraints.