Analysis of temperature distribution, burn-up and breeding parameters in nuclear fuel rod in fusion-fission reactor system fueled with mixed ThO2-UO2 fuel

Abstract

Temperature distribution in nuclear fuel rod, burn-up (BU), possibility of nuclear fuel rejuvenation and breeding parameters are investigated for different coolants under various first wall loads (Pw=2, 5 and 7 MW m−2) in a deuterium–tritium driven fusion–fission reactor system, fueled with mixed UO2-ThO2 fuel. The fuel mixture is considered to be mixed with various mixture fractions. A (D-T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. In the tritium breeding zone of the blanket Li2O is used and blanket is reflected by graphite for neutron economy. The fissile fuel zone is considered to be cooled with three different coolants [gas (He, CO2), Flibe (Li2BeF4) and natural lithium (Li)] with the volumetric ratio of coolant-to-fuel [e=(Vcoolant/Vfuel)=1:1 (45.515% coolant, 45.515% fuel and 8.971% clad)] for nuclear heat transfer. The behavior of the blankets fueled with mixed fuel mentioned above are observed for 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. The fissile fuel breeding occurs through the neutron capture reaction in the 232Th (ThO2), in the 238U (UO2) isotopes. As a result of the calculations, fuel mixtures having the best performance parameters have been obtained as follows for different coolants and under the 7 MW m−2 first wall load during operation period without reaching the melting point: • The blanket fueled with 70 wt% UO2−30 wt% ThO2 for Flibe (Li2BeF4) coolant • The blanket fueled with 80 wt% UO2−20 wt% ThO2 for Gas (He, CO2) coolant • The blanket fueled with 90 wt% UO2−10 wt% ThO2 for natural lithium (Li) coolant Clad surface temperature Tc controlled by coolant velocity is taken between certain operation temperatures (Tc=200, 300, 400, 500 and 600°C). The best Cumulative Fissile Fuel Enrichment (CFFE) grade of the nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials (233U, 235U, 239Pu, and 241Pu) is obtained in Flibe (Li2BeF4) coolant blankets, followed by Gas (He, CO2), whereas natural lithium (Li) coolant shows a poor rejuvenation performance in all fuels. CFFE reaches the maximum value (CFFE=10.6200%) in blanket cooled with Li2BeF4 coolant, followed by gas coolant with 9.0319% and natural lithium coolant with 7.9863% without reaching the melting point (Tmax=Tm>2830°C) of the fuel materials. At this point, the maximum temperature in centerline of the fuel rod (Tm) reach to 2727.57°C in blanket cooled with Li2BeF4 coolant. However, in the blanket fueled with pure UO2 fuel and cooled with Flibe, CFFE and Centerline Temperature of the fuel rods (Tm) are reached to 9.7277% and 2811.13°C respectively at the end of 39 months. In the blanket fueled with pure UO2 and cooled with gas, CFFE and Centerline Temperature of the fuel rods (Tm) are reached to 7.8115% and 2814.57°C at the end of 38 months and 7.2777% and 2724.34°C at the end of the 43 months in the blanket cooled with natural Li. In the blanket fueled with pure ThO2, CFFE reaches the maximum value with 9.2843% in the blanket cooled with Flibe. The enrichment is sufficient for LWR reactor. The temporal change of the fuel composition during the hybrid reactor plant operation and the most pertinent nuclear reactions been considered in the course of BU. The BU reaches the maximum value (72 749.1 MWd t−1 (tone)) in the gas coolant blanket, followed by Flibe coolant blanket with 67 641.0 MWd t−1 natural lithium coolant blanket with 63 897.5 MWd t−1 during the operation period without reaching the melting point. Copyright © 2010 John Wiley & Sons, Ltd.