Transmutation Of Actinides Research Articles

Burnup study on the effective transmutation of minor actinides based on multi-enrichment fuel in the ADS subcritical reactor

In order to solve the problem that the single-enrichment core parameters of the ADS subcritical reactor exceed the safety limit and to improve the MAs effective transmutation in the ADS system, short-term and long-term burnup calculations are carried out based on multi-enrichment fuel and the physical parameters of the sub-critical reactor are calculated and filtered throughout the calculation process. The MAs effective transmutation of the ADS subcritical reactor is calculated using the double-layer nested MAs effective transmutation calculation method. The short-term case is obtained for the target parameters, different fuel partitioning methods, and different fuel composition; long-term burnup calculations are conducted under different refueling methods; finally, a recommended case is given. The ADSOC code, coupled with FLUKA and OpenMC codes, can realize the multi-enrichment fuel refueling, as well as tracking and screening of physical parameters during burnup calculations in the ADS subcritical reactor.

Enhanced transmutation of minor actinides by incorporation of multi-beam design in Accelerator-Driven subcritical system

This study aims to identify the optimal beam design for incinerating the long-lived minor actinides in Accelerator-Driven System (ADS). All the analyses were performed using the MCNP6.1 code with cross-section library ENDF/B-VII. In this study, a depletion method for ADS was conducted and analyses were exercised to investigate the burnup performances for the various arrangements of a three-proton beam design. The influences of the various arrangements of the three-proton beam design were used as a criterion for comparison. It was found that the dual proton beam arrangement demonstrated the best depletion result for the incineration of minor actinides.

Study on damage of Gd2O3-CeO2 under electronic energy loss: comparison between bulk-like and nanostructure.

To understand the physical phenomena responsible for radiation damage of the materials used in nuclear reactors, and thus study their operation life and/or efficiency, it is required to simulate the conditions by exposing the materials to energetic ions. Ceria (CeO2) has been proposed as one of the inert matrices for the transmutation of minor actinides in the futuristic inert matrix fuel (IMF) concept. The inert matrix should also contain burnable poison to compensate for the initial reactivity of fuel. In this context, gadolinium (Gd) is an excellent burnable poison with a high neutron absorption cross-section. In view of this, Gd2O3-CeO2 nano-powders were synthesized and sintered at 800 °C and 1300 °C to obtain different grain sizes and morphologies. FESEM and TEM were carried out to study the grain size of pristine pellets. The sintered pellets were irradiated with 80-MeV Ag ions (electronic energy loss (Se) regime) at room temperature to emulate the effect of fission fragments. For analysis of the effect of grain size on the irradiation-induced structural degradation at different fluences, GIXRD and Raman spectroscopy were performed. Significantly large damage has been observed for the smaller grain-sized samples (sintered at 800 °C) as compared to the large grain-sized sample (sintered at 1300 °C). Neither of the samples amorphized under the present experimental conditions as indicated by the presence of the Raman-active T2g mode (centred at 462 cm-1) and all the XRD peaks of fluorite cubic structure up to the highest fluence employed (1 × 1014 ions cm-2). X-ray photoelectron spectroscopy results demonstrate that Ce4+ to Ce3+ and vacancy-related isolated clusters are the main defects produced in the systems. The radiation tolerance behaviour of the samples is understood with the help of thermal spike simulation, which indicates higher transient lattice temperatures with longer duration in the smaller grain-sized sample upon irradiation. Gd-doped ceria thus possesses good radiation stability in the Se regime, indicating its potential for application in IMFs.

Радиологическая безопасность населения и персонала при двухкомпонентной ядерной энергетике

The Fundamentals of State Policy in the Field of Nuclear and Radiation Safety of the Russian Federation, approved by the President of the Russian Federation, indicate that the goal of state policy in this area is to ensure the protection of the population, considering modern requirements. The IAEA Fundamental Safety Principles emphasize the need to protect both current and future generations from radiation risks. The achievement of these safety goals is ensured by the development in Russia of new nuclear energy system based on a closed nuclear fuel cycle (CNFC) and fast neutron reactors (FNR). One of the objects of the new nuclear energy system is the Experimental Demonstration Energy Complex (EDEC) consisting of the BREST-OD-300 FNR, a reprocessing module and a fuel fabrication/refabrication module, located on the territory of JSC «SHK» in the Tomsk region. This article provides assessments of radiation carcinogenic risks for the population living in the 30-km zone of JSC «SHK» under normal operating conditions and for workers after potential emergency situations at EDEC facilities. The conditions for ensuring the protection of future generations of people are also justified. Radiation risk assessments are obtained directly from the dynamics of equivalent doses in human organs and tissues and modern risk models recommended by ICRP Publication 103. The predicted values of the risks of radiation carcinogenesis for the population and personnel are significantly lower than the current limits of NRB-99/2009 for both normal operation and potential exposure during accidents. It has been shown that transmutation of minor actinides during the development of CNFC based on FNR ensures the safety of future generations of people: it reduces the carcinogenic risk of americium by 213 times, neptunium by 101 times, and curium by 47 times. The effect of radiological equivalence of radioactive waste and natural uranium raw materials is achieved after 99 years of radioactive waste storage. Thus, the priority direction for the development of new nuclear power is the energy sector of the CNFC based on the FNR.

Neutronic evaluation for different external neutron sources in a Small-Subcritical fast reactor

This study proposes an optimized small-sized subcritical fast reactor with an external neutron source for actinide transmutation. The reactor integrates features of Accelerator Driven Systems (ADS) and Fusion-Fission Systems (FFS) based on Tokamak technology. Both external neutron sources were evaluated, one from fusion reactions and the other from spallation reactions in the designed small subcritical fast reactor. The analysis assessed neutron flux and actinide transmutation rates through fuel burnup simulations at 1200 K. The results underscore the significance of external neutron sources in subcritical systems, with the fusion-fission source presenting high actinide transmutation. The SSFR utilizes thorium oxide as fuel and Lead-Bismuth Eutectic (LBE) as coolant. Neutronic parameter comparisons between FFS and ADS neutron sources were conducted, showing advantages for actinide transmutation in the system with the FFS neutron source, which enhances transmutation, minimizing radiotoxicity. The neutronic parameters were obtained using the MCNP, while the MONTEBURNS code, which links the MCNP and the ORIGEN depletion code, was utilized to perform the fuel burnup simulations. The cross-section libraries were generated by the NJOY-2012 using the nuclear data library JEFF-3.3.

A review of alternative finishing options for uranium/plutonium and minor actinide nitrate products from thermal and fast reactor fuels reprocessing

In fuel reprocessing, product finishing is the conversion of aqueous metal nitrates into solid forms that can either be re-used in new fuels or safely interim stored and so is the key step at the interface between reprocessing and fuel manufacturing processes. Conversion processes were originally developed for the fabrication of oxide fuels and have typically involved the generation of UO3 (or U3O8) and PuO2 powders as separate products. However, whilst this is an industrial proven process, research and development of mixed oxide (MOX) fuels and minor actinide targets by advanced reprocessing routes is also underway. It is envisaged that advanced recycle processes will allow the multi-recycling of plutonium and the transmutation of minor actinides into shorter lived isotopes to enhance sustainability of the nuclear fuel cycle and fully utilise fissionable material recoverable from spent fuels whilst improving a number of key features over current conversion processes including: product conversion efficiency, higher throughput, flexibility in product specification, proliferation resistance and a reduction in the number of waste streams produced.Internationally, research programmes to examine future options for advanced fuel cycles are focusing on the development of advanced reprocessing flowsheets for future actinide recycling. It is anticipated that these separation processes will produce a range of mixed transuranic (TRU) actinide nitrate products rather than the separated pure plutonium stream produced in current reprocessing plants.The easiest assumption is that these nitrate products will be converted to oxides by the oxalate co-precipitation route. However, this has certain limitations. The focus of this review is to identify any alternative nitrate to oxide conversion processes which have been applied to mixed oxides and evaluate their suitability for MOX production. A variety of factors including process complexity, technical maturity, effluent treatment/recycling and scale up into an industrial process will also be considered.

An alternative approach for excess reactivity control of TRISO particles

This study explores an innovative approach to manage the initial excess reactivity in High-Temperature Gas-Cooled Reactors (HTGRs) by modifying the fuel and non-fuel compositions of TRISO particles. Choosing a single fuel block of the 350 MWth Prismatic HTR as the reference, minor actinides (MAs) in three loading amounts (2 wt.%, 5 wt.%, and 10 wt.%) were introduced into the TRISO fuel kernels across three alternating rings of the assembly. Four more models were developed by replacing the TRISO particles in these rings with QUADRISO, featuring distinct layers of four burnable absorbers (B4C, Er2O3, Eu2O3, and Gd2O3), each having a thickness of 2 µm. These approaches were compared against using conventional B4C poison rods, placed at adjacent and opposite ends of the hexagonal assembly. Depletion analysis over a period of 600 EFPDs was carried out using open-source Monte Carlo code OpenMC. The Eu2O3 layered particles and oppositely placed B4C rods exhibited the most effective reduction in initial reactivity, amounting to 0.28 Δk/k and 0.17 Δk/k respectively. The 10 wt.% MA loading approach proved to be versatile, as it efficiently lowered the initial reactivity by 0.15 Δk/k while also achieving minor actinide transmutation, yielding transmutation rates of 23.1%/yr for 237Np, 51.5%/yr for 241Am, and 30%/yr for 243Am. Fuel cycle parameters were calculated for multi-batch refueling schemes using the linear reactivity model (LRM). While MAs reduced burnup and cycle length, the use of B4C, Er2O3 and Eu2O3 layers improved fuel cycle performance, with B4C demonstrating the best results. Due to its high absorption cross section, Gd2O3 rendered the core sub-critical initially, resulting in negative fuel cycle parameters. The fuel temperature coefficient (FTC), relative pin-power distribution, and energy-dependent neutron flux of the models were also evaluated.

Preparation and thermal physical properties of MgO-(Nd1-xYx)2(Zr1-xCex)2O7 composite ceramics as a candidate for inert matrix fuel

In recent years, the MgO-Nd2Zr2O7 composite ceramics inert matrix (IM) was proposed as a candidate for the transmutation of Pu and minor actinides (MA) in light water reactors (LWR) or accelerator driven sub-critical system (ADS). To take full advantage of the pyrochlore structure of Nd2Zr2O7 to accommodate different types of MA and Pu, in this work, Y and Ce co-doped ωMgO-(1-ω)(Nd1-xYx)2(Zr1-xCex)2O7 (M-NYZC) composite ceramics used as a candidate IM for inert matrix fuel were prepared by solid-state one-step sintering method at 1500 °C for 3 h. The phase structure and microstructure of M-NYZC composite ceramics were characterized by XRD and SEM-EDS. Also, the thermal physical properties of M-NYZC composite ceramics including the thermal expansion coefficient and thermal conductivity were investigated systematically. Results from XRD and SEM-EDS showed that the magnesium oxide and NYZC pyrochlore solid solution are intermixed well with each other, and the M-NYZC samples present uniform element distribution with a grain size of about 1 μm. Also, the as-prepared samples possessed well densified microstructure, and the average relative density of all samples could exceed 95 %. Moreover, the thermal expansion coefficient and thermal conductivity of 0.5M-0.5NYZC (x = 0, 0.15, 0.2, 0.5, 0.9) samples were located in the range of 13.4 × 10-6–14.4 × 10-6 K-1 and 16.34–3.70 W·m-1·K-1 respectively. By comparison, the thermal physical properties of 0.5M-0.5NYZC samples are superior to that of traditional UO2 and MOX fuels. It is suggested that the M-NYZC composite ceramics prepared in this study can meet the thermal physical performances of inert matrix fuel.

Enhancing transmutation rates of minor actinides using Zr and Y hydride and deuteride coatings in an LFR

The possibility of transmutation of minor actinides (MAs) in lead-cooled fast reactors provides an excellent opportunity to reduce the overall repository of radioactive waste. The present work seeks to augment the transmutation rates of six minor actinides (237Np, 241Am, 243Am, 243Cm, 244Cm, and 245Cm) by increasing neutron capture through the use of moderator coatings around the fuel elements. The reference design for the fuel assembly was based on the design concept of ALFRED, the European Lead Cooled Demonstrator Reactor, and consists of 127 fuel elements loaded with 5 wt.% of minor actinides (MAs) homogeneously mixed with MOX (mixed oxide) fuel. Four different moderators (ZrH1.6, ZrD1.6, YH2, and YD2) with five different coating thicknesses (0.01 cm, 0.02 cm, 0.03 cm, 0.04 cm, and 0.05 cm) were selected for this study and variation in kinf, transmutation rates and fuel cycle parameters (employing the linear reactivity model or LRM) were determined. Results showed that both the addition of a moderator and the increase in moderator thickness improved transmutation rates but had a negative impact on fuel cycle parameters. ZrH1.6 resulted in the highest increase in transmutation rate, but also incurred the largest penalties in fuel burnup, and cycle length. For ZrH1.6 coatings of 0.01 cm, the obtained transmutation rates were 13.57%/y for 237Np, 12.6%/y for 241Am, and 9.26%/y for 243Am. The concentration of Cm isotopes increased over time due to neutron capture of Am, U, and Pu. Deuteride moderators minimized the decrease in fuel cycle parameters but compromised transmutation rates. The fuel temperature coefficient (FTC), energy dependent neutron flux, and relative pin power distribution of the reference design and the models were also determined in order to analyze the effect of moderator addition on these factors. Finally, a suggestion of the best model was put forth in consideration of both transmutation efficiency and fuel cycle parameters.

Feasibility Study for Minor Actinides Transmutation in Conventional Power Reactors

The handling and storage of the spent fuel from the current nuclear power reactors is one of the major ecological challenges in front of the nuclear energy industry, due to its long period of high radiotoxicity. It is mainly caused by the long-lived minor actinides generated in the nuclear fuel. Therefore profound study of the possibilities for shortening the terms of storage of spent nuclear fuels and some of high radioactive waste through the currently available technologies is necessary. The possibility for transmutation of Neptunium, Americium and Curium isotopes, as most abundant in the spent nuclear fuel of the wide-spread Light Water Reactors is evaluated in the paper using one of the most recent tools, available for the purpose. The most technologically accessible option, the current nuclear power reactors is analysed. The fuel in the cores of the LWR (PWR) and the PHWR (CANDU) type of reactors are modelled and the variation of the quantity of Neptunium, Americium and Curium isotopes after typical fuel campaigns for these reactors is estimated. The obtained results show that under the investigated conditions both types of reactors show potential for decrease of the minor actinides quantity in spent fuel before its final disposal, therefore provide a vital option toward sustainable and environmentally friendly nuclear spent fuel management.

Calculated demonstration of VVER-1000 SNF actinide transmutation efficiency

Calculated demonstration of VVER-1000 SNF actinide transmutation efficiency

Reliability analysis of core power optimization control using Kalman filter for accelerator driven system based on reconfigurable computing

Accelerator Driven System (ADS) is a promising nuclear facility for the transmutation of minor actinides and long-lived fission products, which is constructed by proton accelerator, spallation target and sub-critical reactor. For the industrial implementation, the reliability of core power control system in the operation process has to be analyzed and verified in considering the dynamic characteristics and real time response for the safety operation strategy. Therefore, a digitalization of Instrumentation and Control (I&C) system for ADS that can predict the changing value of proton beam intensity has been suggested in this research to get the appropriate variation of reactor core power without arousing the unexpected impulse to the main structures. In order to accurately analyze the random behavior of the precursor concentrations and the neutron density in sub-critical reactor, the stochastic differential equation model has been developed, and the closed-loop control system in describing the mapping relationship between the proton beam intensity and the corresponding core power has been established. The Kalman filter has been adopted for the optimal design of proposed analysis method to achieve the competitive predicting performance in operation process, and all the algorithms have been implemented on the XILINX ZYNQ System-on-Chip (SoC) using reconfigurable computing technology to meet the real time design requirements for embedded applications. Finally, the reliability of core power control model has been tested against Monte Carlo calculations and all the results demonstrate the good performance in predicting the core power with acceptable efficiency.

Investigation on Neutronic Behavior of Pebble Bed Reactor for TRU Transmutation

One alternative effort to eliminate the accumulation of transuranic (TRU) elements generated from LWR spent fuel is by utilizing TRU as nuclear fuel in pebble bed reactor. This work was aimed to investigate the neutronic behavior of a pebble bed reactor used for TRU transmutation. The reactor geometry is adopted from the HTR-Modul. MCNP6 multipurpose radiation transport code with ENDF/B-VII.1 neutron library was used to calculate the neutronic aspects. The results indicated that the initial effective multiplication factor (keff) values decrease as the amount of TRU fuel pebbles within the core increases, concurrent with the fuel burnup. Increasing the TRU fuel pebble ratio resulted in weakened Doppler temperature coefficients (DTCs) and improvement in moderator temperature coefficients (MTCs). The calculated total mass of plutonium and minor actinide transmutation demonstrated that a pebble bed reactor with 100% TRU pebble can reduce TRU mass up to 55.47% of its initial mass load.

How colloid nature drives the interactions between actinide and carboxylic surfactant in sol: Towards a mesostructured nanoporous actinide oxide material

HypothesisThe key to prepare a mesostructured porous material by a soft-template route coupled to a colloidal sol–gel process is to control the surfactant-colloid interface. In the case of tetravalent actinide ions, their high reactivity in aqueous media always leads to uncontrolled and irreversible condensation. The addition of a complexing agent to the sol may moderate these reactions and enhances the interaction between the colloids and the surfactant to in fine prepare a mesostructured nanoporous actinide oxide material. ExperimentsSeveral colloidal sols were prepared without and with formic acid as complexing agent by varying the molar ratios between thorium, carboxylic surfactant and pH. Small and Wide Angle X-ray Scattering were used to characterize the nature of the colloids, their interaction with the surfactant and the final ThO2 materials. FindingsDepending on the colloid nature, hexagonal or worm-like hybrid mesophase is formed. The thermal treatment of the worm-like mesophase with a sufficient amount of Th-formic acid hexameric species coated at the surface of surfactant micelles generates micrometric ThO2 nanofibers. This material having an accessible porosity opens new perspectives to be impregnated with minor actinide solutions offering a promising safety method for the fabrication of mixed oxide nuclear fuel and the minor actinide transmutation.

Radiation response of Y3Al5O12 and Nd3+-Y3Al5O12 to Swift heavy ions: insight into structural damage and defect dynamics.

Understanding the behavior of a material under irradiation is paramount to its application in the nuclear industry. The present work explores the radiation response of garnet Y3Al5O12 (YAG) and Nd3+-substituted Y3Al5O12 (Nd-YAG) under a 100 MeV Iodine beam at varying fluences to mimic the effect of fission fragments. This is relevant to the potential application of garnet as a host for minor actinide (MA) transmutation (Nd3+: surrogate for long-lived MA (Am3+, Np3+, Cm3+)). The un-irradiated and irradiated YAG and Nd-YAG samples were investigated by X-ray diffraction and Raman spectroscopy. Positron annihilation spectroscopy, thermal spike modelling and theoretical studies have been employed to understand the role of substitution and defect energetics in influencing this radiation response. Although both materials were not completely amorphized under the present irradiation conditions, a tremendous loss in crystallinity could be observed with increase in fluence, the damage being much more in Nd-YAG. Ion track radii of 2.17 nm and 2.91 nm were estimated for YAG and Nd-YAG respectively. Thermal-spike calculations show an increase in radiation-induced transient temperatures upon Nd-substitution that causes greater radiation damage in Nd-YAG. The enhancement in radiation-induced damage with increasing ion-fluence manifests in broadening and weakening of the Raman modes and XRD peaks. An increase in the average positron annihilation lifetime indicated the creation of oxygen vacancies. The defect formation energies of Y3Al5O12 have been theoretically estimated via density functional theory (DFT) and unfavorable energies required for creating cation pair anti-sites have been proposed as one of the possible reasons for the relatively poorer radiation response of YAG. The irradiation behavior of Y3Al5O12 has been compared with disordered fluorite (YSZ) and zirconate pyrochlores, which are well-researched ceramics for MA transmutation.

Multi-group analysis of Minor Actinides transmutation in a Fusion Hybrid Reactor

New nuclear technologies are currently being study to face High Level Waste treatment and disposal issues. Generally, GEN-IV fission Fast Reactors (FR) are considered the waste-burners of the future. In fact, a fast flux turns out to be the best choice for actinides irradiation in critical reactors because of favorable cross section conditions. Differently, Fusion Fission Hybrid Reactors (FFHR) are futuristic devices based on the combination of fusion and fission systems and could represent an alternative to FRs. In such systems, the choice spectrum of the neutron flux that irradiates HLW may be non-obvious due to some operational constraints which have to be considered. To design and optimize these systems as waste-burners, one should fully understand the transmutation dynamics occurring into the fission region. A multi-energy-group analysis by FISPACT-II code has been set to analyze the conversion processes in scenarios characterized by different neutron energy spectra and fluences. The results of this study show that, despite fast fluxes are characterized by better behaviors in terms of radiotoxicity treatment, the difficulties of reaching high reaction yields may require solutions involving moderators or broadened neutron fluxes to increase the reactions probabilities and, consequently, actinides mass conversion yield.

Viscosity of fluoride melts promising for molten salt nuclear reactors

The viscosity of molten salt, as an important hydrodynamic property, should be taken into account when creating and operating molten salt nuclear reactors (MSRs). An eutectic FLiNaK is considered to be one of the most suitable for use in MSR designed for the minor actinides transmutation. The dynamic viscosity of the molten mixtures FLiNaK + NdF3, FLiNaK + CeF3 and FLiNaK + LaF3 was measured in a temperature range of 600–700 °C using the high-temperature rotary rheometer FRS-1600. Lanthanide fluorides were considered as analogues of actinide fluorides. It was revealed that the additions of rare earth fluorides (REM)F3 in amount of 15 mol. % significantly impact the viscosity of the system FLiNaK + (REM)F3,but the effect of NdF3, CeF3 and LaF3 was found to be almost the same. In order to calculate the kinematic viscosity of the molten mixture FLiNaK + NdF3, a regression equation depending on several parameters was derived. This model equation can be used for predicting the kinematic viscosity of molten mixtures of FLiNaK with other rare earth fluorides.

Minor Actinides Transmutation and 233U Breeding in a Closed Th-U Cycle Based on Molten Chloride Salt Fast Reactor

Long-lived minor actinides (MAs) are one of the primary contributors to the long-term radiological hazards of nuclear waste, and the buildup of MAs is hampering the development of nuclear power. The transmutation of MAs in reactors is regarded as a potential way to replace direct disposal to reduce the impact of MA on the environment and improve the utilization of fuel. Due to its superior features, such as outstanding neutron economy, no fuel assembly fabrication, high neutron flux, and especially online refueling and reprocessing, the molten chloride salt fast reactor (MCFR) is regarded as one of the potential reactors for MA incineration. In this work, MA transmutation capability and 233U breeding performance for an optimized MCFR have been evaluated in different scenarios. The results show that the MA transmutation capability and 233U breeding performance with online transuranic elements (TRU) and 232Th feeding scenario are improved significantly compared with the case in online 233U and 232Th feeding, when the initial MA loading is 5 mol%, the total mass of MA transmutation and MA incineration is 7160 kg and 1759 kg during the whole 100 years operation under online TRU and 232Th feeding scenario, and the corresponding average annual net production of 233U is 450 kg, however, the MA transmutation amount, MA incineration amount and average annual net production of 233U for online 233U and 232Th feeding scenario is 5298 kg, 1315 kg, and 249 kg, respectively. In addition, the research also shows that the increase in initial loading of MA has no obvious effect on the improvement of the 233U breeding performance but can improve the transmutation efficiency of MA under online TRU and 232Th feeding scenarios. Furthermore, if 233U is continuously extracted online from the core during the operation, the 233U breeding performance will be significantly improved, but it will deteriorate the safety performance, such as the fuel temperature coefficient of reactivity (TCR) and the effective delayed neutron fraction (EDNF), more importantly, it will also put forward higher requirements for the immature online reprocessing technology.

Physical feasibility of minor actinides transmutation in a two-component nuclear energy system in Russia

A transition to a two-component nuclear power structure with a reactor fleet consisting of thermal and fast reactors as envisioned in the Russian nuclear power development strategy to 2050 and outlook to 2100 will require optimal spent nuclear fuel and radioactive waste management solutions. A core issue in this regard is managing the long-lived minor actinide (MA) inventory that affects overall nuclear power ecological safety. The study examines several options for homogenous MA (Am and Np) transmutation using modern calculation codes with MA transmutation rate and material balances taken into account. Results demonstrate that if fast reactor installed capacity reaches 92 GWe by 2100 there would not be any need for dedicated MA-burners as the MA issue would be gradually resolved within the two-component nuclear energy system by the end of the century.

Minor Actinides Transmutation Performance in a Closed Th–U Cycle Based on Molten Chloride Salt Fast Reactor

The molten chloride salt fast reactor (MCFR) with a closed Th-U fuel cycle is receiving more and more attention due to its excellent performance, such as high solubility of actinides, superior breeding capacity, and good inherent safety. In this work, the neutronics performances for different minor actinides (MA) loadings and operation modes are analyzed and discussed based on an optimized MCFR. The results indicate that online continuous reprocessing can significantly increase the transmutation performance of MAs. In addition, MA loadings have an obvious effect on the neutronics characteristics of the MCFR, and it is helpful for improving the MA transmutation capability and 233U breeding performance, simultaneously. When MA = 5 mol%, the average annual MA transmutation mass and incineration mass can achieve about 53 kg and 13 kg, respectively, and the corresponding annual net production of 233U is 250 kg. When MA = 33.5 mol%, the annual MA transmutation mass and incineration mass can be about 310 kg and 77 kg, respectively, and the corresponding annual net production of 233U is 349 kg. However, when the MA loadings exceed 10%, the corresponding keff will exceed 1.1 for decades, even if only Th is continuously fed online. The results also indicate that the transmutation ratio (TR) and incineration ratio (IR) of MA increase and reach maximum values in the first decades for all the different MA loadings, which means MA may be fed into the fuel salt to improve its transmutation capability. Moreover, though MA loading will increase the level of radiotoxicity of the core in the early stage of burnup, the radiotoxicity of MA will drop rapidly after a brief rise during the operation. It can also be found that the temperature coefficient of reactivity (TCR) of all different MA loadings can be negative enough to maintain the safety of the MCFR during the whole operation, although it decreases in the beginning of life (BOL) with the increasing MA loading. Furthermore, the evolution of an effective delayed neutron fraction (EDNF) is also researched and discussed, and the EDNF varies most significantly when loading MA = 35.5 mol%, with a range of 273 to 310 pcm over the entire 100 years of operation.