Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

Abstract

Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th) O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library

“…In case of fueling the core with UO 2 , β eff values were 0.0078 and 0.00688 with ENDF/B-VII.1 and ENDF/B-VIII, respectively, while the expected value from the 235 U fission is 0.00650 according to Lamarsh and Baratta [23], so it was realistic to see the core has a delayed neutron fraction that is greater than this expected value. According to RFANE [19], the delayed neutron fraction is 0.00740 and 0.007110 according to Gholamzadeh et al [20] which is in agreement with the obtained results.…”

“…e neutronic parameters were calculated by MCNP6 with ENDF/B-VII.1 and ENDF/B-VIII and compared with the reactor design parameters [19] and published results [9,20]. It is necessary for any reactor core to have delayed neutrons since they can control the increase in the reactor power [21].…”

Heat transfer in a simulated spent fuel pool for different types of spent fuels

“…Another kinds of mixed fuels have been investigating e.g., oxides of thorium/uranium, thorium/plutonium, thorium/transuranic, and transuranic/uranium [1,[3][4][5][6]. Recently, researchers have been turning their attention to Th fuel cycle in PWRs aiming at reducing the generation of minor actinides, at improving the nuclear power sustainability, and at better fuel utilization and breeding [7][8][9]. Studies have also discussed the possible role of Th utilization in power reactors and the associated fuel cycles [10].…”