RESUMO
A comprehensive analysis has been performed to investigate the conversion of the Syrian MNSR (miniature neutron source reactor) from current HEU fuel to selected alternatives LEU and MEU fuels. For this purposes the core design calculations related to design and engineering of LEU and MEU fuels have been carried out using the codes WIMSD/4 and BORGES-part of the MTR-PC and the code CITATION. Aiming at reducing the fuel enrichment by maintaining reactor power, thermal neutron flux and excess reactivity in the same range of the current MNSR design, two fuel alternatives of LEU (UO(2)-Mg) and MEU (U(3)Si(x)-Al) have been investigated. The results indicate that the first type (UO(2)-Mg) realizes the criticality conditions with low enrichment of 20% using the similar overall design of the present HEU fuel pins, whereas the second type (U(3)Si-Al) requires increasing the enrichment up to 33%. For the purpose of reactor core lifetime extension the possibility of mixing the burnable poisons Gd(157) and Cd(113) in the fresh core has been also explored. Thus, the calculation results indicate that the long-term control effect of Cd(113) on the excess reactivity is more homogeneous over the time due to the lower burn up rate of this burnable poison.
RESUMO
The codes WIMSD/4 and BORGES--part of the MTR-PC code package--have been applied to prepare the microscopic cross-section library for the main elements of miniature neutron source reactor (MNSR) core for six neutron energy groups. The generated library has been utilized by the 3D code CITATION to perform the calculation of fuel burn-up including the identification of main fission products and their impacts on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products. The burn-up results have indicated that the core life-time of MNSR is being mainly estimated by Sm(149) followed by Gd(157) and Cd(113). The accumulation of these fission products during 100 continuous operation days caused a reduction of about 4.3 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 3.5 mk, which relates to the whole discontinuous operation period of the reactor since its start up to now. The calculation procedure simulates the sporadic operation with an equivalent continuous operation period. This approximation is valid for the long-lived fission products that mainly dictate the core life-time. However, it is an overestimation for the concentration of short-lived radioactive products like Xe(135).
