Thorium abundance in the Earth's crust is estimated to be three to four times higher than uranium. This is one potential advantage of Thorium as a provider of attractive fuel to produce nuclear energy. Fewer transuranics produced by Thorium during the fuel burn up in the reactor may also be another advantage for reducing the long-term burden of high-level long-lived waste. The scope of this paper is to study the implication of Thorium fraction on neutronic parameters of pebble bed reactor. The reactor model of HTR-10 was selected, and the (Th, 235U)O2 fuel was used in this study. The MCNP6 code was applied to solve a series of neutron transport calculations with various Thorium fractions in (Th,235U)O2 fuel based on the ENDF/BVII library. The calculation results show that the total temperature coefficient of reactivity of Thorium-added pebble bed reactors is generally more negative than those of LEU-fuelled one, except for 10% Thorium fraction. The kinetic parameters, especially prompt neutron lifetime and neutron generation time of pebble bed reactors, are higher, which means the addition of Thorium in the fuel makes the reactor more easily controlled. However, the burn-up calculations show that the introduction of Thorium in the same fuel kernel as LEU within the pebble bed reactor is unable to lengthen the fuel residence time, except for a minimum of 40% Thorium fraction.
Molten salt reactor (MSR) is often associated with thorium fuel cycle, thanks to its excellent neutron economy and online reprocessing capability. However, since 233U, the fissile used in pure thorium fuel cycle, is not commercially available, the MSR must be started with other fissile nuclides. Different fissile yields different inherent safety characteristics, and thus must be assessed accordingly. This paper investigates the inherent safety aspects of one fluid MSR (OF-MSR) using various fissile fuel, namely low-enriched uranium (LEU), reactor grade plutonium (RGPu), and reactor grade plutonium + minor actinides (PuMA). The calculation was performed using MCNPX2.6.0 programme with ENDF/B-VII library. Parameters assessed are temperature coefficient of reactivity (TCR) and void coefficient of reactivity (VCR). The result shows that TCR for LEU, RGPu, and PuMA are -3.13 pcm, -2.02 pcm and -1.79 pcm, respectively. Meanwhile, the VCR is negative only for LEU, whilst RGPu and PuMA suffer from positive void reactivity. Therefore, for the OF-MSR design used in this study, LEU is the only safe option as OF-MSR starting fuel.Keywords: MSR, Temperature coefficient of reactivity, Void coefficient of reactivity, Low enriched uranium, Reactor grade plutonium, Minor actinides
Thorium-based nuclear fuel has become an interesting subject for a variety of research with a wide range of applications. Research focusing on thorium-based fuel is aimed to overcome the scarcity and limitation of natural uranium resources as an alternative nuclear fuel in a thermal reactor. As thorium has no naturally occurring fissile isotope, it requires other fissile isotopes in order to be converted into fissile 233U to produce energy. The isotopes 235U and 2359Pu are two of the few alternatives available as the fissile nuclei for a thorium-fueled reactor. The purpose of this paper is to investigate the criticality and burnup performance of pebble bed reactor using two options of thorium-based fuel–namely, UO2-ThO2 and PuO2-ThO2. The HTR-10 was chosen as the reactor model. A series of criticality calculations with various uranium contents in UO2-ThO2 fuel and various plutonium contents in PuO2-ThO2 fuel was conducted using the Monte Carlo transport code MCNP6 and continuous energy nuclear data library ENDF/B-VII. The calculation result shows that 35% plutonium content in PuO2-ThO2 fuel has comparable criticality with 80% uranium content in UO2-ThO2 fuel. The former is shown to be better to ensure longer reactor operational time. However, that is due to the higher fissile fraction compared to that of the latter fuel. Meanwhile, thorium played little part in prolonging reactor criticality, as 233U production is not particularly significant.
also be considered for the simplicity it offers and thus potentially lower cost. Due to different neutronic and burnup profile between the two, the resulting spent fuel characteristic is also different and possibly requires different handling mechanism. This paper assesses the characteristics of OTTO-scheme RDE spent fuel using thorium fuel cycle to provide preliminary data and insight for its spent fuel management. The assessment is performed employing ORIGEN2.1 code. At day 30 of cooling after determined end-of-cycle (EOC), each spent fuel yields 234.9 Curies of radioactivity, emitting 66.26 neutrons/second, 1x10 13 photons/second, and releasing 0.7675 watts of decay heat. These numbers must be taken into consideration regarding spent fuel management and spent fuel cask design. Tl-208 isotope characteristics, whose existence is unique to thorium fuel cycle, were also determined. It is found to be yielding 3.42x10-3 Curie of radioactivity and releasing 1.2x10 8 photons/second at its peak. Understanding its high-energy gamma release, proper radiation protection mechanism must be implemented.
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