Transmutation Of Minor Actinides Research Articles

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.

Facile Preparation of Macro-Microporous Thorium Oxide via a Colloidal Sol-Gel Route toward Safe MOX Fuel Fabrication.

The identification of new colloidal sol-gel routes for the preparation of actinide oxides, which have a homogeneous and accessible porosity that can easily be impregnated by any concentrated actinide solution, opens new perspectives for the preparation of homogeneous nuclear fuel for minor actinide transmutation. This homogeneity allows us to avoid "hot spot" formation due to the local accumulation of more fissile elements. Here, we report the preparation of macro-microporous ThO2 materials by a colloidal sol-gel route. Using a thorium salt with 6-aminocaproic acid as a complexing agent at a controlled pH, we were able to pilot the condensation of thorium hydroxo species forming colloids of tuned nanometric size and thus the sol stability. After a freeze-drying process to concentrate colloids and a thermal treatment allowing complexing agent removal and macroporosity formation by a brutal gas release during combustion, a loose packing of ThO2 nanoparticles with an ordered distribution of interparticular porosity and a fraction of nanometric crystallites, whose size depends on the initial colloidal size, were obtained. The sols, pastes, and final materials were characterized by small- and wide-angle X-ray scattering to determine the colloidal size and the final structure of the materials, which was also confirmed by transmission electron microscopy. The most promising material was finally successfully impregnated by a simulating minor actinide solution and thermally treated to prepare a mixed actinide oxide material. This safe technology, relying on the colloidal sol-gel process and the formulation of complex fluids forming tunable precursors, opens new perspectives for the reuse of nuclear waste solutions as new fuel.

Beam physics design of a 30-MW beam transport to the target for an accelerator-driven subcritical system

The Japan Atomic Energy Agency's accelerator-driven subcritical system (JAEA-ADS) proposes the reduction of nuclear waste through the transmutation of minor actinides. The JAEA-ADS drives a 30-MW proton beam to a spallation target to produce neutrons for a subcritical 800-MWth reactor. The beam must be transported from the end of the linear accelerator (linac) to the target located inside the reactor core with high beam power stability and low peak density to ensure beam window integrity, which is a primary concern for the ADS project. Additionally, the design of the beam transport to the target (BTT) must be compatible with the established reactor design, and the elements that comprise the BTT must facilitate the maintenance and replacement of the fuel and the beam window. To this end, a robust-compact BTT design was developed through massive multiparticle simulations. First, the beam optics was optimized to guarantee beam window feasibility requirements by providing a low peak density of less than 0.3 μA/mm2. Then, beam stability was evaluated and improved by a simultaneous application of input beam and element errors. The input beam errors were based on the beam degradation obtained by implementing fast fault compensation in the linac, which is a key strategy to attain high-reliability operation. The results show that the BTT design fulfills reactor and beam window requirements for JAEA-ADS.

Beam dynamics studies for fast beam trip recovery of the Japan Atomic Energy Agency accelerator-driven subcritical system

High reliability and availability are primary goals for the operation of particle accelerators, especially for accelerator-driven subcritical systems (ADS). ADSs employ high-power beams for the transmutation of minor actinide; as a result, the amount and the radiotoxicity of the nuclear waste are considerably reduced. To this end, the Japan Atomic Energy Agency is designing a 30-MW continuous wave (cw) superconducting proton linear accelerator (linac) that supplies neutrons to an 800-MW subcritical reactor by a spallation process. The major challenge for an ADS linac is the strict control of the beam trip duration and its frequency to avoid thermal stress in the subcritical reactor structures. The maximum allowed beam trips for failures longer than a few seconds are estimated to be far below the rate achieved in current accelerators. Thus, we implemented a combination of hot standby and local compensation that enables a fast beam recovery. This work comprehensively investigated the tolerance of our linac lattice for the local compensations for failures in superconducting cavities and magnets. This scheme includes simultaneous compensation of multiple cavities in independent and same cryomodules that significantly enhance the reliability of the linac. The retuned schemes present acceptable beam performance to guarantee the integrity of the linac and the beam transport to the target; moreover, they satisfy the beam stability in the beam window. In addition, the readjusted elements are subjected to moderate stress to ensure a sustainable operation. This manuscript reports the beam dynamics results toward fulfilling the high reliability demanded by an ADS linac.

Study Analysis of Multiplication Factor on ADS Transmutation System

New research technology for partition and transmutation (P-T) of minor actinide (MA) contained in high-level waste (HLW) has been carried out to design accelerator-based transmutation system or commonly known as Accelerator Driven Subcritical (ADS) device. The objective of this system is to eliminate or minimize high-level radioactive waste that has long half-life and to develop long-term safety assurance in HLW management. Analysis done by using Origen2.1 and mCnP6 codes. As matrix, Th-232 with optimum weight 75% is used.. Comparing with , analysis results using MCNP shows that the use of PWR nuclear spent fuel for ADS device can be done by reprocessing to eliminate U-238 nuclide, which is the source of the formation of plutonium nuclide and minor actinides. From the results and discussion, it can be concluded that observation on transmutation of transuranic nuclides using ORIGEN2.1 Code on ADS device has been conducted successfully with LMFBR library. 15%-Th232 fuel has been justified as the most optimum ADS fuel based on K-inf reached at EOC on various fuel types. Mass reduction occurs at EOC for U-236, U-235, U-234, Th-232, Np-237, Pu-238, Am-241, Pu-242, and Cm 244. Radionuclides that experience mass changes during cooling time are Pu-238, Pu-240, Am-241, Pa-233, and Cm-244.

Preparation of MgO-Nd2Zr2O7 composite ceramics used for inert matrix fuel by one-step sintering method

MgO-Nd2Zr2O7 composite ceramic is a promising candidate as an inert matrix material for the transmutation of Pu and minor actinide. In this work, the MgO-Nd2Zr2O7 composite ceramics (hereafter M-NZO) with 30–70 wt% mass percent of MgO were successfully prepared at 1500 °C for 24 h via direct solid-state reaction one-step method. This method achieved the integration of Nd2Zr2O7 pyrochlore synthesis and M-NZO composites sintering, which shows a significant improvement in the simplification of the preparation process than the traditional two-step method. The structural properties and compactness of the as-prepared composite ceramics were systematically investigated by X-ray diffraction (XRD), scanning electron microscopy (SEM). It was confirmed that the as-prepared M-NZO ceramics are composed of only crystalline MgO and Nd2Zr2O7 phases, and the mass fractions of two phase are almost the same as the designed nominal chemical composition. The composite ceramics showed remarkable densification and the relative density is higher than 98.2%.

Comparison of the minor actinide transmutation efficiency in models of a fast neutron uranium-thorium fueled reactor

In terms of nuclear raw materials, the issue of involving thorium in the fuel cycle is hardly very relevant. However, in view of the large-scale nuclear power development, the use of thorium seems to be quite natural and reasonable. The substitution of traditional uranium-plutonium fuel for uranium-thorium fuel in fast neutron reactors will significantly reduce the production of minor actinides, which will make it attractive for the transmutation of long-lived radioactive isotopes of americium, curium and neptunium that have already been and are still being accumulated. Due to the absence of uranium-233 in nature, the use of thorium in the nuclear power industry requires a closed fuel cycle. At the initial stage of developing the uranium-thorium cycle, it is proposed to use uranium-235 instead of uranium-233 as nuclear fuel. Studies have been carried out on the transmutation of minor actinides in a fast neutron reactor in which the uranium-thorium cycle is implemented. Several options for the structure of the core of such a reactor have been considered. It has been shown that heterogeneous placement of americium leads to higher rates of its transmutation than homogeneous placement does.

Nuclear transmutation by flux compression

Abstract A new idea for the transmutation of minor actinides, long (and even short) lived fission products is presented. It is based an the property of neutron flux compression in nuclear (fast and/or thermal) reactors possessing spatially non-stationary critical masses. An advantage factor for the burn-up fluence of the elements to be transmuted in the order of magnitude of 100 and more is obtainable compared with the classical way of transmutation. Three typical examples of such transmuters (a subcritical ringreactor with a rotating reflector, a sub-critical ring reactor with a rotating spallation source, the socalled “pulsed energy amplifier”, and a fast burn-wave reactor) are presented and analysed with regard to this purpose.

Minor actinides transmutation in pressurized water reactors. 1. Multiple recycling of minor actinides on the example of one VVER reactor

This article explores the possibilities and conditions of combustion in a pressurized water reactor of its own accumulated minor actinides (MA). The simplest computational model is used: an infinitely extended medium with the distribution and composition of all materials of the fuel assembly of the reactor core, similar to VVER-1200, with uranium dioxide having an initial 235U enrichment of 4.95%. The burnup model is presented in the form of iterations, each of which simulates a fuel campaign lasting 4 years without refueling. At the start of the cycle, special fuel rods are loaded with minor actinides extracted from the reprocessed SNF of the VVER-1200 reactor. After the end of the fuel campaign, all the MAs are removed from the SNF and used in a new iteration. As a result of calculations, it was found that the MA mass in the cycle after 3–7 iterations (depending on the number of fuel elements allocated for the placement and accumulation of MAs) tends to an equilibrium state (regardless of the MAs added every 4 years). In other words, the fuel rods allocated for loading MAs play the role of a kind of furnace, into which, in each iteration, MAs from the previous iteration accumulated in the given reactor are loaded. After several iterations, the burned MA mass converted into fission products is compared with the incoming one. The inclusion of MAs in this way into the fuel cycle converts at least 86% of MAs into fission products without affecting the power generation of the nuclear power plant. It is important that MAs are temporarily unloaded from the reactor after the next iteration in order to remove fission products and to add a new portion of MAs. After stopping the reactor operation, about 16% of the total amount of MAs generated for the entire history of the reactor’s life is discharged into the storage facility. The initial fuel composition in the fuel rods allocated for loading MAs differs from the others only in the amount of MAs and the mass of 238U. The simplified computational model used in this work (without annual overloads of the reactor) influenced the burnup depth and, naturally, the duration of operation, i.e., the k∞ value becomes less than 1 after 1056 days instead of the actual 1460 days with annual fuel overloads. This affected the average fuel composition and, consequently, the neutron spectrum, and could affect the main result of the work, i.e., the number of burned-out MAs in different iterations. Additional calculations, taking into account the annual overloads of the reactor, showed that the change in the spectral composition had little effect on the amount of MAs at the end of the fuel campaign (within 2%). It turned out that the replacement of 238U with minor actinides in fuel rods, the number of which is less than 10, leads to a loss of reactivity. When the number of fuel rods for loading MAs is more than 10, the reactivity increases, giving hope for burning up MAs accumulated in several reactors.

A design study on a metal fuel fast reactor core for high efficiency minor actinide transmutation by loading silicon carbide composite material

ABSTRACT A metal fuel fast reactor core for high efficiency minor actinide (MA) transmutation was designed by loading silicon carbide composite material (SiC/SiC) which can improve sodium-cooled fast reactor (SFR) core safety characteristics such as sodium void reactivity worth and Doppler coefficient due to neutron moderation. Based on a 750 MWe metal fueled SFR core concept designed in a prior work, the reactor core loading fuel subassemblies with SiC/SiC wrapper tubes and moderator subassemblies was designed. To improve the reactor core safety characteristics efficiently, three layers of SiC/SiC moderator subassemblies were loaded in the core by replacing 108 out of 393 fuel subassemblies with the moderators. The reactor core with approximately 20 wt% MA-containing metal fuel satisfied all safety design criteria and achieved the MA transmutation amount as high as 420 kg/GWe-y which is twice as high as that of the axially heterogeneous core with inner blanket and upper sodium plenum, and two-thirds of that of the accelerator-driven system.

Fuel depletion characteristics of MA transmutation in PWR

The minor actinide transmutation in PWR can change its fuel depletion characteristics distinctly. We simulate PWR reactor fuel depletion characteristics of minor actinide different admitting approaches on burnable poison rods for transmutation. Simulation results indicate that minor actinide introduction in water gaps of burnable poison rods and minor actinide replacement of burnable poison in the burnable poison rods cause some unfavorable impacts to the core lifetime, whereas minor actinide mixing with burnable poison shortens the core lifetime significantly. For the same minor actinide introduction amount, the impact to the fuel cycles of a thin layer minor actinide coating on PWR burnable poison rods is the least, whereas the impact to the fuel cycles of minor actinide mixing with burnable poison homogeneously is the greatest.When implementing minor actinide transmutation on PWR one needs to balance the effects on PWR fuel cycles and the reactivity coefficients, and then selects an appropriate minor actinide introduction approach on the PWR reactor core.

Assessment of radiation damage of the first wall of a fusion neutron source DEMO-FNS with a blanket for the transmutation of minor actinides

The design of the demonstration fusion neutron source (DEMO-FNS) demonstration fusion–fission hybrid facility is carried out at the National Research Center ‘Kurchatov Institute’ in Russia. For the implementation of the DEMO-FNS project, it is important to assess the effect of the fast neutron spectrum on radiation-induced damage to the device materials, primarily the materials of the first wall, which is the most problematic unit of the device. Due to the lack of real hybrid fusion devices, the study can only be carried out using computer simulations of the experiment. A computer simulation using the Monte Carlo method was carried out to calculate the developed three-dimensional full-scale model of the DEMO-FNS reactor with a blanket for the transmutation of minor actinides. The MCNP-4 code was used with cross-sections from the FENDL-2.1 and ENDF/B-6 files, as well as with cross sections for calculating radiation displacements. The neutron spectrum in the first wall of DEMO-FNS was determined at the power of a volumetric deuterium–tritium neutron of 1.42 × 1019 n s−1. It was found that beryllium as a plasma-facing material has too short lifetime under the neutron load on the first wall and divertor plates of DEMO-FNS. The replacement beryllium to tungsten is problematic and requires special study. Copper alloys have sufficient resistance to the effects of neutron radiation of DEMO-FNS for one full power year, but will require several replacements when operating the device for more than 10 years. The steels used in fast reactors can meet the operational requirements of the first wall of DEMO-FNS. To solve the problem of choosing materials for the first wall of DEMO-FNS, new experimental researches on changes in the physical properties of these materials in the fast neutron spectrum are needed.

SUPERFACT: A Model Fuel for Studying the Evolution of the Microstructure of Spent Nuclear Fuel during Storage/Disposal.

The transmutation of minor actinides (in particular, Np and Am), which are among the main contributors to spent fuel α-radiotoxicity, was studied in the SUPERFACT irradiation. Several types of transmutation UO2-based fuels were produced, differing by their minor actinide content (241Am, 237Np, Pu), and irradiated in the Phénix fast reactor. Due to the high content in rather short-lived alpha-decaying actinides, both the archive, but also the irradiated fuels, cumulated an alpha dose during a laboratory time scale, which is comparable to that of standard LWR fuels during centuries/millenaries of storage. Transmission Electron Microscopy was performed to assess the evolution of the microstructure of the SUPERFACT archive and irradiated fuel. This was compared to conventional irradiated spent fuel (i.e., after years of storage) and to other 238Pu-doped UO2 for which the equivalent storage time would span over centuries. It could be shown that the microstructure of these fluorites does not degrade significantly from low to very high alpha-damage doses, and that helium bubbles precipitate.

Study on transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions

The feasibility of transmutation of minor actinides recycled from the spent nuclear fuel in the VVER-1000 LEU (low enriched uranium) fuel assembly as burnable poison was examined in our previous study. However, only the minor actinide vector of the VVER-440 spent fuel was considered. In this paper, various vectors of minor actinides recycled from the spent fuel of VVER-440, PWR-1000, and VVER-1000 reactors were therefore employed in the analysis in order to investigate the minor actinide transmutation efficiency of the VVER-1000 fuel assembly with different minor actinide compositions. The comparative analysis was conducted for the two models of minor actinide loading in the LEU fuel assembly: homogeneous mixing in the UGD (Uranium-Gadolinium) pins and coating a thin layer to the UGD pins. The parameters to be analysed and compared include the reactivity of the LEU fuel assembly versus burnup and the transmutation of minor actinide nuclides when loading different minor actinide vectors into the LEU fuel assembly.

Ab Initio molecular dynamics study of threshold displacement energy in Zirconium Nitride

• The threshold displacement energy ( E d ) in ZrN was evaluated by the AIMD method. • The weighted average values of E d were 33 eV and 29 eV for Zr and N atoms, respectively, from the several crystallographic directions investigated in this study. • The values of E d and the collision process were significantly dependent on the type of atom and crystallographic direction. • The [110] direction exhibited the minimum value of E d , which is associated with sequential replacement collisions. • The preferable configurations for Zr and N interstitials were the <111> split-type and the tetrahedral sites, respectively, which influence the difference in the values of E d . Zirconium nitride (ZrN) is a promising matrix candidate for advanced nuclear fuels and transmutation of minor actinides. This study investigates the displacement process induced by low-energy recoils in ZrN using ab initio molecular dynamics (AIMD) simulations to evaluate the threshold displacement energy ( E d ). Observations of the collision processes of primary knock-on atoms (PKAs) for both Zr and N atoms were performed for seven different directions: [100], [110], [111], [210], [211], [221], and [321], which cover most regions of the stereographic triangle. The values of E d ranged from 15 eV to 50 eV, and the collision processes were dependent on the crystallographic orientations. The weighted average values of E d evaluated from the PKA directions investigated in this study were 33 eV and 29 eV for the Zr and N atoms, respectively. Anti-site defects were not formed for either Zr or N PKAs. Sequential replacement collisions along the [110] atomic row played an important role in the collision process, providing a lower value of E d in the crystallographic directions. Furthermore, configurations of interstitials were different between the Zr and N atoms, which also influenced E d values.

Current state of partitioning and transmutation studies for advanced nuclear fuel cycles

This paper aims at giving a general overview of the current state of partitioning and transmutation (P&T) studies for advanced nuclear fuel cycles across the world. P&T usually refers to the entire fuel cycle steps linked with minor actinides transmutation, from separation of the minor actinides from the spent fuel to manufacturing of new fuels suited for transmutation, irradiation, handling and reprocessing or disposal of the spent transmutation fuels. The feasibility of most of these steps has been demonstrated in the laboratory scale since the beginning of the nuclear era, however the physical-chemical and radiological properties of minor actinides make their handling considerably more complicated than standard UOX or MOX fuel. As such, research on this topic is still ongoing around the world.Due to the shutdown of various experimental installations since the 90′ (EBR-II in the US in 1994, PFR in the UK in the same year, Joyo in Japan in 2007, Phénix and Osiris in France in 2009 and 2016 respectively), few irradiation experiments have been carried out in the past decade. The MARIOS and DIAMINO experiments were carried out in the HFR and OSIRIS reactor and showed that the irradiation of fuel with a micro-structure optimized for helium release (and thus limited swelling) was achievable. The irradiation of metallic and oxide containing minor actinides and/or rare earths was carried out in the ATR reactor in the US. The good behaviour of oxide fuel containing minor actinides was demonstrated, although pin failures occurred with the metallic fuel. Furthermore, Post-Irradiation Examinations of various irradiation experiments have been carried out, the most salient one being the METAPHIX experiment in Phenix, which demonstrated the safe behaviour under irradiation of the base alloy considered in the experiment.Nevertheless, encouraging results in the field of americium separation from the spent fuel have been obtained, the main one being the successful demonstration of the ExAm process to isolate americium from a PUREX raffinate. Furthermore, progress has been made on the manufacturing of UAmO2 targets during the preparation of DIAMINO and MARIOS experiments. Finally, interest in minor actinides transmutation has started to rise again, notably linked with the current trend in molten salt reactor, which are suitable for minor actinides transmutation.

A methodology for evaluating the transmutation efficiency of long-lived minor actinides

Up to now no definite internationally recognized quantitative criterion of minor actinides (MAs) transmutation efficiency was worked out, although this would be highly desirable. The absolute and relative total mass reduction of MAs are completely inadequate because they ignore the accumulation of higher radiotoxic long-lived MAs from the transmuted nuclide. In the current work, we introduce a new criterion for transmutation efficiency of MAs in nuclear reactors and demonstrate its efficiency by comparing two molten salt reactors; the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and the Small Molten Salt Fast Reactor (SMSFR). Our proposed criterion takes into account the mass of all useful actinides, short-lived MAs, and short-lived fission products (FPs). In contrast, the mass parameters calculate the reduction in the MAs mass regardless of the produced nuclides. We introduce a new approach to load MAs into both reactors. The proposed approach merges the advantages of both homogeneous and heterogeneous approaches. The overall change in the actinides and FPs mass during the irradiation has been calculated using direct SERPENT-2 calculations. The results show that the transmutation efficiency of 241Am (the prime isotope for the transmutation) in the SD-TMSR is much higher than in the SMSFR. After 1500 days of radiation, the transmutation efficiency reaches 82.6% for SD-TMSR, however, for SMSFR it reaches 52.5%.