Transmutation Of Minor Actinides Research Articles

Transmuting minor actinides with thermal reactor neutrons

The closure of uranium-based fuel cycle assumes that Uranium and Plutonium are extracted from the spent fuel. Among remaining heavy nuclides, the most radioactive are minor actinides. One of the concepts to reduce the reprocessed fuel radioactivity is the repeated long-term irradiation of minor actinides (this process is referred to as transmutation). There are works stating the necessity of hard spectrum neutrons for irradiation, for which special reactors-transmuters and accelerator-driven systems are to be included into the fuel cycle.In this work, practicability of minor actinide transmutation in thermal spectrum reactors is considered. The main factor representing the transmutation practicability that is used in this work, is the ratio of radioacitivities with transmutation, and without, ξ(t). This ratio as a function of time can be greater or less than 1, and its values define to the most extent the transmutation practicability. We computed functions ξ(t) for minor actinides. The conclusions are (from the reduction of radioactivity point of view): Neptunium transmutation is hardly necessary, since the order of magnitude reduction in radioactivity is first reached in hundred thousand years; rather small effect (the order of magnitude radioactivity reduction in 500 years) is found for Curium; the Americium radioactivity is reduced by factor of 10–100 in 300 years after irradiation. It is shown that the best values of ξ(t) are reached at 70–08% burnup of minor actinides.The final conclusion about the practicability of Americium and Curium transmutation must be drawn by taking into account in the considered scenarios the difference in probability of the environmental release, the difference of biological effect and the transmutation efficiency of minor actinides continuously fed to spent fuel storages by the operating nuclear energy industry.

Thoria based inert matrix fuel production via sol–gel process

There is an important concern that uranium reserves will not be sufficient for new nuclear facilities in the future. Thorium could offer a potential alternative fuel because it is three times more abundant than uranium. Thorium based nuclear fuels are important candidates for the future of nuclear fuels. The main interest is in considering thorium based fuels for burning plutonium in conventional reactors. Thoria based inert matrix fuels (IMF) are also candidates for burning plutonium and the transmutation of minor actinides. In this study, inert matrix fuel production was examined by external gelation process. Metal nitrate solution of each material (thorium nitrate, cerium nitrate, aluminum nitrate and magnesium nitrate) were mixed for the feed solution. The metal nitrate solution was 50% of total feed solution. After mixing the metal nitrate, solution polyvinyl alcohol (PVA, 10%) and tetrahydrofurfuryl alcohol (THFA, 40%) were added into mixed metal nitrate solution during stirring. PVA, THFA and metal nitrate solution were dropped through a nozzle into gelation medium ammonia solution. Then external gelation process was carried out. Cerium (Ce) was used to simulate Plutonium (Pu). Obtained gels from the external gelation process were aged, washed and dried. The powders were calcinated at 1273 K for 6 h. After calcination, the powders were pressed at 500 MPa with a manual press. Pellets were sintered at elevated temperatures (1473, 1573, 1673, 1773, 1873, and 1973 K) for 4 h. The densities of sintered pellets were measured by immersion method. Sintered pellets characteristics were determined by scanning electron microscopy, x-ray diffraction analysis and energy dispersive x-ray analysis and the results were discussed.

Minor actinide transmutation in fast reactor metal fuels irradiated for 120 and 360 equivalent full-power days

An irradiation experiment on uranium–plutonium–zirconium (U–Pu–Zr) alloys containing 5 wt% or less minor actinides (MAs) and rare earths was carried out in the Phénix fast reactor. The isotope compositions of the fuel alloys irradiated for 120 and 360 equivalent full-power days (EFPDs) were chemically analyzed by inductively coupled plasma–mass spectrometry after 3.3–5.3 years of cooling. The results of chemical analysis indicated that the discharged burnups of the fuel alloys irradiated for 120 and 360 EFPDs were 2.1–2.5 and 5.3–6.4 at%, respectively. The changes in the isotopic abundances of plutonium, americium, and curium during the irradiation experiment were assessed to discuss the transmutation performance of MA nuclides added to U–Pu–Zr alloy fuel. Multigroup three-dimensional diffusion and burnup calculations accurately predicted the changes in these isotopic abundances after fuel fabrication. An evaluation of the MA transmutation ratio based on the results of chemical analysis revealed that the quantity of MA elements in the U–19Pu–10Zr–5MA (wt%) alloy decreased by about 20% during the irradiation experiment for 360 EFPDs.

Prediction of MgO volume fraction in an ADS fresh fuel for the scenario code CLASS

Subcritical reactors, also called Accelerator Driven Systems (ADS), are specifically studied for their capacity in transmuting Minor Actinides (MA). Nuclear fuel cycle scenarios involving MA transmutation in ADS are widely researched. The nuclear fuel cycle simulation tool code CLASS (Core Library for Advanced Scenarios Simulations) is dedicated to the inventory evolution calculation induced by a complex nuclear fleet. For managing reactors, the code CLASS includes physic models. Loading models aim to provide the fuel composition at beginning of cycle according to the stocks isotopic composition and the reactors requirements. A cross section predictor aims to provide mean cross sections needed for solving Bateman equations. Physic models are built from reactors calculation set ahead of the scenario calculation. An ADS standard composition at BOC is a mixture of plutonium and MA oxide. The high number of fissile isotopes present in the subcritical core leads to an issue for building an ADS fuel loading model. A high number of isotopic vector at BOC is needed to get an exhaustive simulation set. Also, ADS initial reactivity is adjusted with an inert matrix which induces an additional degree of freedom. The building of an ADS fuel loading model for CLASS requires two steps. For any heavy nuclide composition at beginning of cycle, the core reactivity must be imposed at a subcritical level. Also, the reactivity coefficient evolution should be maintained during the irradiation. In this work, the MgO volume fraction is adjusted to reach the first requirement. The methodology based on a set of reactor simulations and neural network utilization to predict the MgO volume fraction needed to reach a wanted keff for any initial composition is presented. Also, a complete neutronic study is done that highlight the effect on MgO on neutronic parameters. Reactor simulations are done with the transport code MCNP6 (Monte Carlo N particle transport code). The ADS geometry is based on the EFIT (European Facility for Industrial-Scale Transmutation) concept. The simulation set is composed of more than 8000 randomized runs from which a neural network has been built. The resulting MgO prediction method allows reaching a keff at 0.96 and the distribution standard deviation is around 200 pcm.

Analysis of minor actinides transmutation for a Molten Salt Fast Reactor

As one of the six candidate reactors chosen by the Generation IV International Forum (GIF), Molten Salt Fast Reactor (MSFR) has many outstanding advantages and features for advanced nuclear fuel utilization. Effective transmutation of minor actinides (MA) could be attained in this kind of fast reactor, which is of importance in the future closed nuclear fuel cycle scenario. In this work, we attempt to study the MA transmutation capability in a MSFR with power of 500MWth by analyzing the neutronics characteristics for different MA loadings. The calculated results show that MA loading plays an important role in the reactivity evolution of the MSFR. A larger MA loading is favorable to improving the MA transmutation performance and simultaneously to reducing the fissile consumption. When MA=18.17mol%, the transmutation fraction can achieve to about 95% on iso-breeding. We also find that although the fuel temperature coefficient (FTC) decreases with the increasing MA loading, it is still negative enough to keep the safety of the MSFR during the whole operation time. The MA contribution to the effective delayed neutron fraction (EDNF) and the intensity of spontaneous fission neutron (ISFN) are also analyzed. Also MA loading can affect the EDNF during the operation and the ISFN of the MSFR is dominated by 244Cm. Finally, we analyze the effect of the core power on MA transmutation capability. The result shows that for all the operating powers the depletion ratio of MA to HN increases with time and reaches a maximum value. And additional MA should be fed into the fuel salt before the MA depletion ratio reaches the peak value to improve its transmutation capability. The net mass of the transmuted MA during the 50years operation for 500MWth is 5620kg which is very close to that of 1000MWth.

ASTRID-like Fast Reactor Cores for Burning Plutonium and Minor Actinides

Abstract A reduction of nuclear waste by transmutation of trans-uranium elements (TRUs), such as Pu and Minor Actinides (MAs) contained in Spent Nuclear Fuel (SNF), is a goal for future reactors. In general, countries with on-going nuclear scenarios would profit from MA mass stabilization, while transmutation of Pu and MAs from SNF could be desired in countries in nuclear phase-out. Both missions can be accomplished by employing fast reactors loaded with fuels containing different amounts of Pu and MAs in a closed (or partially closed) fuel cycle. In this paper, two 1200 MWth sodium-cooled fast reactor cores, based on the French ASTRID design, are proposed for burning TRUs (phase-out option) or only MAs (on-going option). Main attention is focused on the safety and on the transmutation performance. The coolant void effect, in the region including the core and the plenum above and the Doppler constant of both systems are negative also with irradiated fuel. The conversion ratios (CR) of the Pu and MA burners are in the ranges from 0.6 to 0.7 and from 0.5 to 0.6, respectively. These results show a large safety and transmutation potential of ASTRID type reactors.

Study of reactivity control method by the application of gadolinium hydride to accelerator-driven system

A reactivity control method for accelerator-driven system (ADS) is studied for its ability to reduce both the maximum beam current and the load on the beam window. A burnable position (BP) assembly (with gadolinium and zirconium hydride pins) is applied to the ADS core for reactivity control, and various BP assembly optimizations (such as pin arrangement, BP assembly loading position, and BP composition) are performed to minimize the burnup reactivity swing. These optimizations lead to a decreased burnup reactivity swing that is as much as 82.5% less than the swing of non-BP-loaded core; furthermore, the maximum beam current is 12.5 mA. The reactor characteristics of the optimized BP-loaded ADS core are also analyzed to investigate how the introduction of BP assembly influences the system. Safety parameters (such as the Doppler coefficient and the coolant void reactivity) worsen with the introduction of BP assembly, and the total minor actinide transmutation amount decreases 30–40 kg because of the moderated neutrons and changed fuel composition.

MYRRHA: Preliminary front-end engineering design

The MYRRHA project started in 1998 by SCK•CEN. MYRRHA is an MTR, based on the ADS concept, for material and fuel research, for studying the feasibility of transmutation of Minor Actinides and Long-Lived Fission Products arising from radioactive waste reprocessing and finally for demonstrating at a reasonable power scale the principle of the ADS. The MYRRHA design has progressed through various framework programmes of the European Commission in the context of Partitioning and Transmutation. The design has now entered into the Front End Engineering Phase (FEED) covering the period 2012–2015. The engineering company, which will handle this phase, has been selected and the works have begun in the late 2013. In the mean time we have made some refinements in both primary systems and plant layout, including the reactor building design. In this paper, we present the most recent developments of the MYRRHA design as existing today.

Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0wt.%, 4.0wt.%, 6.0wt.%, 8.0wt.%, 10.0wt.% and 12.0wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library.In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void worth increased with the MA weight fraction, while the Doppler feedback and the effective delayed neutron fraction decreased with the increase of the MA weight fraction.

Concept of DT Fuel Cycle for a Fusion Neutron Source

A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 1018 neutrons per second is under designe in Russia. Such a FNS is of interest for many applications: (i) basic and applied research (neutron scattering, etc); (ii) testing the structural materials for fusion reactors; (iii) control of sub-critical nuclear systems and (iv) nuclear waste processing (including transmutation of minor actinides). This paper describes of fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ~0.5 and ~0.3m, magnetic field ~1.5 T, heating power less than 15MW and plasma current 1–2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems.

Sensitivity analysis of minor actinides transmutation to physical and technological parameters

Minor actinides transmutation is one of the three main axis defined by the 2006 French law for management of nuclear waste, along with long-term storage and use of a deep geological repository. Transmutation options for critical systems can be divided in two different approaches: (a) homogeneous transmutation, in which minor actinides are mixed with the fuel. This exhibits the drawback of “polluting” the entire fuel cycle with minor actinides and also has an important impact on core reactivity coefficients such as Doppler Effect or sodium void worth for fast reactors when the minor actinides fraction increases above 3 to 5% depending on the core; (b) heterogeneous transmutation, in which minor actinides are inserted into transmutation targets which can be located in the center or in the periphery of the core. This presents the advantage of decoupling the management of the minor actinides from the conventional fuel and not impacting the core reactivity coefficients. In both cases, the design and analyses of potential transmutation systems have been carried out in the frame of Gen IV fast reactor using a “perturbation” approach in which nominal power reactor parameters are modified to accommodate the loading of minor actinides. However, when designing such a transmutation strategy, parameters from all steps of the fuel cycle must be taken into account, such as spent fuel heat load, gamma or neutron sources or fabrication feasibility. Considering a multi-recycling strategy of minor actinides, an analysis of relevant estimators necessary to fully analyze a transmutation strategy has been performed in this work and a sensitivity analysis of these estimators to a broad choice of reactors and fuel cycle parameters has been carried out. No threshold or percolation effects were observed. Saturation of transmutation rate with regards to several parameters has been observed, namely the minor actinides volume fraction and the irradiation time. Estimators of interest that have been derived from this approach include the maximum neutron source and decay heat load acceptable at reprocessing and fabrication steps, which influence among other things the total minor actinides inventory, the overall complexity of the cycle and the size of the geological repository. Based on this analysis, a new methodology to assess transmutation strategies is proposed.

Simulation of an Accelerator Driven Subcritical Core with Mixed Uranium-Thorium Fuel

During recent years, a new generation of nuclear reactors, known as “Accelerator Driven Subcritical Reactors”, has been developed. One of the new application aspects for such reactors (besides transmutation of High Level Waste and burning Minor Actinides) is usage of thorium as nuclear fuel. In this work a subcritical core in experimental scale is simulated by MCNPX code. The core contains two types of fuel assemblies: (85% ThO2 + 15% UO2) and MOX (U-Pu). In the first step, only the thorium-contained fuel assemblies are loaded into the core. Criticality calculations using MCNPX show that the keff is so low that the fuel assemblies cannot run the subcritical core. This implies that MOX (U-Pu) assemblies must be loaded as well. Neutronic parameters of the thorium- fueled Accelerator Driven Subcritical core are then calculated as well as some other parameters related to accelerator coupled with the core. The main objective of this simulation is to study the behavior of Accelerator Driven Subcritical core with thorium assemblies.

The MYRRHA ADS Project in Belgium Enters the Front End Engineering Phase

The MYRRHA project started in 1998 by SCK•CEN. MYRRHA is a MTR, based on the ADS concept, for material and fuel research, for studying the feasibility of transmutation of Minor Actinides and Long-Lived Fission Products arising from radioactive waste reprocessing and finally for demonstrating at a reasonable power scale the principle of the ADS. The MYRRHA design has progressed through various framework programmes of the European Commission in the context of Partitioning and Transmutation. The design has now entered into the Front End Engineering Phase (FEED) covering the period 2012-2015. The engineering company, which will handle this phase, has been selected and the works have begun in the late 2013. In the mean time we have made some refinements in both primary systems and plant layout, including reactor building design. In this paper, we present the most recent developments of the MYRRHA design in terms of reactor building and plant layout as existing today as well as a preliminary study concerning the spent fuel building of the facility. During the oral presentation we add some preliminary results of the interaction with the FEED contractor and the most recent version of the primary systems.

Minor actinide transmutation on PWR burnable poison rods

Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core.We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR keff markedly. The PWR keff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons.Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and composition of nuclear fuel. Furthermore, the two minor actinide loading patterns completely isolate minor actinide transmutation materials from nuclear fuel, so it also facilitates fuel reprocessing after discharged from PWR core. We believe that loading minor actinide to PWR burnable poison rods for transmutation is an optimal minor actinide loading pattern.

Pyrochemical treatment of spent nitride fuels for MA transmutation

Nitride fuels have several advantages including high thermal conductivity and high metal density (like metallic fuels) and high melting point and isotropic crystal structure (like oxide fuels). Since the late 1990s, the partitioning and transmutation of minor actinides (MA) has been studied to decrease the long-term radio-toxicity of high-level waste and to mitigate the burden of final disposal. Japan Atomic Energy Agency (JAEA) has proposed a dedicated transmutation cycle using an accelerator-driven system (ADS) with nitride fuels containing MA. The nitride fuel cycle we have developed includes a pyrochemical process. Our focus is on the electrolysis of nitride fuels and their refabrication from the recovered actinides; other processes are similar to the technology for metal fuel treatment and have been studied elsewhere. Here, we summarize our activity on the development of the pyrochemical treatment of spent nitride fuels.

Design of J-PARC Transmutation Experimental Facility

After the Fukushima accident, nuclear transmutation got much interested as an effective option of nuclear waste management. Japan Atomic Energy Agency (JAEA) proposes the transmutation of minor actinides by accelerator-driven system (ADS) using lead–bismuth eutectic alloy (Pb–Bi) as a spallation target and a coolant of subcritical core. Current design of ADS has 800 MWth of rated power, which is driven by 20 MW proton LINAC, to transmute minor actinides generated from 10 units of standard Light Water Reactors. To obtain the data required for ADS design, JAEA plans to build a Transmutation Experimental Facility (TEF) in J-PARC. TEF consists of two buildings, one is an ADS target test facility (TEF-T), which will be installed Pb–Bi spallation target, and the other is Transmutation Physics Experimental Facility (TEF-P), which set up a fast critical/subcritical assembly. TEF will be located at the end of 400 MeV LINAC of J-PARC and accepts 250 kW proton beam. The paper describes an outline of TEF and design of spallation target.

Uncertainty analysis of infinite homogeneous lead and sodium cooled fast reactors at beginning of life

The objective of the present work is to estimate breeding ratio, radiation damage rate and minor actinide transmutation rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed taking into account uncertainty in nuclear data and composition of the reactors. We use the recently released ENDF/B-VII.1 nuclear data library and restrict the work to the beginning of reactor life.We work under multigroup approximation. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. Modeling error and numerical error are estimated.The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the responses. The generalized adjoint fluxes are used to calculate first order sensitivities of the responses to model parameters. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the responses.We show that the uncertainty in model parameters is the dominant source of uncertainty, followed by modeling error, input data precision and numerical error. The uncertainty due to composition of the reactor is low. We identify main sources of uncertainty and note that the low-fidelity evaluation of 16O is problematic due to lack of correlation between total and elastic reactions.

Improved multi-group cross sections with resonance treatment for future applications involving minor actinides

In the future study for thorium reactors and minor actinide transmutation reactors, resonance self-shielding treatment for them becomes remarkable issue for criticality and isotope depletion. Resonance treatment for important minor actinides and other resonant absorbers has been carried out with subgroup method. The updated multi-group cross sections and subgroup data is replaced with existing multi-group library. The results from neutron transport code nTRACER with updated multi-group cross section library were compared with existing library and solutions from MCNP. The resonance interaction of uranium with important minor actinides has been accommodated through modified interference treatment by interference correction in subgroup method. The results for cross sections and multiplication factor for pin and assembly problems showed improvement from systematic resonance treatment for resonant absorber nuclides including 237Np and 243Am.

Technology readiness assessment of partitioning and transmutation in Japan and issues toward closed fuel cycle

This paper treats a technology readiness assessment of partitioning and transmutation in Japan and issues toward closed fuel cycles. The generic technology readiness level in this study is based on the definition in the Global Nuclear Energy Partnership: TRL 3 shows the status that critical function are proved and elemental technologies are identified; TRL 4 represents the level that the related technologies are validated at bench-scale in laboratory environment; and TRL 5 indicates the completion of the development related to the subsystem and elemental technologies. The reviewed technological area includes the partitioning and transmutation technologies for minor actinide cycle: fast breeder reactor, and accelerator driven system for minor actinide transmutation; partitioning processes; and minor actinide bearing fuels. The assessments reveal that the TRLs stay around the final step of concept development (TRL 3) and the first half step of elemental technology development (TRL 4) because each system requires more development of its elemental technologies. The fast breeder reactor is assessed to be at TRL 4 because critical experiments with americium and neptunium would be additionally required, and the technology for the accelerator driven system reaches TRL 3 due to several researches. Consequently, a common key issue is how nuclear calculation methodology will be validated for the MA-bearing-fuelled core; however, critical experiments with several kilogrammes of americium or more are difficult in the existing experimental facilities. On the other hand, engineering scale tests of the MA partitioning processes using actual spent fuel, and engineering scale of fabrication and irradiation tests with the separated materials are required to achieve the second half step of elemental technology development (TRL 5); however, they could be a massive investment. Therefore, laboratory-scale tests using actual material are proposed. The tests simulate a nuclear fuel cycle: partitioning actual spent fuels; and fabricating MA bearing fuels from the extracted materials; and then irradiating them. It should be that the tests advance technological readiness from TRL 4 to TRL 4+, which is a reasonable and feasible pathway.

Analysis of advanced European nuclear fuel cycle scenarios including transmutation and economic estimates

Four European fuel cycle scenarios involving transmutation options (in coherence with PATEROS and CP-ESFR EU projects) have been addressed from a point of view of resources utilization and economic estimates. Scenarios include: (i) the current fleet using Light Water Reactor (LWR) technology and open fuel cycle, (ii) full replacement of the initial fleet with Fast Reactors (FR) burning U–Pu MOX fuel, (iii) closed fuel cycle with Minor Actinide (MA) transmutation in a fraction of the FR fleet, and (iv) closed fuel cycle with MA transmutation in dedicated Accelerator Driven Systems (ADS). All scenarios consider an intermediate period of GEN-III+ LWR deployment and they extend for 200years, looking for long term equilibrium mass flow achievement.The simulations were made using the TR_EVOL code, capable to assess the management of the nuclear mass streams in the scenario as well as economics for the estimation of the levelized cost of electricity (LCOE) and other costs.Results reveal that all scenarios are feasible according to nuclear resources demand (natural and depleted U, and Pu). Additionally, we have found as expected that the FR scenario reduces considerably the Pu inventory in repositories compared to the reference scenario. The elimination of the LWR MA legacy requires a maximum of 55% fraction (i.e., a peak value of 44 FR units) of the FR fleet dedicated to transmutation (MA in MOX fuel, homogeneous transmutation) or an average of 28 units of ADS plants (i.e., a peak value of 51 ADS units).Regarding the economic analysis, the main usefulness of the provided economic results is for relative comparison of scenarios and breakdown of LCOE contributors rather than provision of absolute values, as technological readiness levels are low for most of the advanced fuel cycle stages. The obtained estimations show an increase of LCOE – averaged over the whole period – with respect to the reference open cycle scenario of 20% for Pu management scenario and around 35% for both transmutation scenarios. The main contribution to LCOE is the capital costs of new facilities, quantified between 60% and 69% depending on the scenario. An uncertainty analysis is provided around assumed low and high values of processes and technologies.