Advanced Nuclear Reactors—Challenges Related to the Reprocessing of Spent Nuclear Fuel

Abstract

Nuclear energy can help stop climate change by generating large amounts of emission-free electricity. Nuclear reactor designs are continually being developed to be more fuel efficient, safer, easier to construct, and to produce less nuclear waste. The term advanced nuclear reactors refers either to Generation III+ and Generation IV or small modular reactors. Every reactor is associated with the nuclear fuel cycle that must be economically viable and competitive. An important matter is optimization of fissile materials used in reactor and/or reprocessing of spent fuel and reuse. Currently operating reactors use the open cycle or partially closed cycle. Generation IV reactors are intended to play a significant role in reaching a fully closed cycle. At the same time, we can observe the growing interest in development of small modular reactors worldwide. SMRs can adopt either fuel cycle; they can be flexible depending on their design and fuel type. Spent nuclear fuel management should be an integral part of the development of new reactors. The proper management methods of the radioactive waste and spent fuel should be considered at an early stage of construction. The aim of this paper is to highlight the challenges related to reprocessing of new forms of nuclear fuel.

1. Introduction

The term advanced nuclear reactors refers either to Generation III+ and Generation IV or small modular reactors. Every nuclear reactor type has an associated fuel cycle. The nuclear fuel cycle is the series of industrial processes that involve the production of electricity from uranium in nuclear power reactors [1]. The nuclear fuel cycle consists of two phases: the front-end and the back-end. The front-end steps prepare uranium for use in nuclear reactors. The back-end of the nuclear fuel cycle includes all processes and activities related to the management of spent nuclear fuel (SNF) after it has been removed from a reactor. SNF is a highly radioactive material and must be handled and stored with special precautions due to its potential hazards. It needs to be stored for hundreds of thousands of years, until its radioactivity naturally decays to safe levels, comparable to naturally occurring uranium ore [2]. SNF contains a significant amount of usable fissile material (uranium and plutonium) that can be separated and then reused as fuel in reactors. The reprocessing and reusing Pu and U as MOX (mixed oxide fuel) fuel is currently realized by closed-cycle nuclear, as shown in Figure 1. In 2024, about 30% of spent fuel had been reprocessed, mostly in France (La Hague) [3]. It is worth noting that for the French nuclear industry, recycling spent fuel is a legal obligation. Recycling spent fuel supports the secure supply of uranium, which is consistent with the country’s strategic interest in meeting its demands through its own mining. Currently, mining accounts for about 15% of total demand, with recycling providing an additional 25%, but the long-term goal is to meet 40% of demand by reprocessing [4]. The reprocessing plants also operate in Russia, Ozersk (Mayak), Japan (Rokkasho), the UK (Sellafield), and India (four plants). However, there are limits to the number of times nuclear fuel can be recycled in thermal reactors due to the changing isotopic composition of the plutonium during each cycle. Fast reactors can address this challenge and efficiently burn certain types of nuclear waste, also including minor actinide [5]. The third strategy presented in Figure 1, fully closed nuclear fuel cycle is not currently used anywhere in the world but it is under intensive development to make nuclear energy more sustainable in the future. The commercialization will require advanced fuel cycle technologies (e.g., advanced fuel fabrication and reprocessing) and significant initial investment in fast reactor infrastructure.

Waste management principles focus on minimizing waste and maximizing resource recovery [7]. The concept is based on a waste hierarchy that prioritizes prevention (avoid), then reduction (minimization), reuse and recycling, and finally disposal (Figure 2). The rules also apply to radioactive waste and are called “integrated cradle to grave waste management”. It states that planning for activities should begin before radioactive waste is generated. It is key to ensuring the future sustainability of nuclear energy.

It is necessary to note that the size of reactors does not influence the amount of generated waste. Kim et al. [8] compared selected parameters, among them volume and mass of SNF for different SMRs and PWR as a reference conventional reactor. It was shown that the amount of waste generated per unit of energy from SMRs and conventional reactors is similar.

2. Spent Fuel Reprocessing—State of the Art

The reprocessing of spent uranium fuels has been investigated since the 1950s [3]. The most common method is based on the dissolution of spent nuclear fuel in concentrated nitric acid and then the solvent–solvent extraction of actinides using an extracting agent in the organic solvent (Figure 3) [9]. The widespread extracting agent is tributyl phosphate used in Plutonium Uranium Reduction Extraction (PUREX) [10]. PUREX is the only process that is currently successfully used on a commercial scale [11].

Uranium and Plutonium are separated from fission products and minor actinides. This is a key process for producing MOX (mixed UO₂ and PuO₂) fuel used in thermal reactors and therefore in a partially closed cycle. Obtaining pure plutonium products increases the risk of proliferation. Because of that, the adjustments of the PUREX process to obtain a direct mix Pu-U were made, resulting in, e.g., the COEX process in France [12], the Advanced PUREX process used in Japan [13], and the UREX process in the USA [14]. The lanthanide fission products and the minor actinides (neptunium, americium, curium, etc.) remain in the aqueous raffinate that is the high-level waste (HLW).

The separation process will play a crucial role in the advanced nuclear fuel cycle [15]. The general conception is that the fissile materials are repeatedly recycled into fuels for advanced reactors such as Generation IV fast reactors to maximize the energy value of the fuel components. This strategy is also called plutonium multi-recycling in Fast Breeder Reactors. The chemical similarity of minor actinides and the lanthanide fission products makes separation complicated. The current research focuses on selective minor actinides multi-recycle for transmutation that will allow a fully closed fuel cycle. Many extractant molecules were tested, differing in the strength of their interaction with trivalent actinides and lanthanides [16,17]. Some lanthanide radioisotopes are so-called “reactor poisons”. They are characterized by high neutron capture cross-sections compared to the minor actinides. To ensure efficient transmutation without the production of new toxic, long-lived radionuclides, it is necessary to obtain pure fractions of plutonium and americium [18]. Specifically, the extraction of the 3+ oxidation state of americium poses a challenge, which was described in many review papers [19,20,21,22,23].

Baron et al. [11] reviewed all processes developed to separate various elements from spent fuel. The article discussed both hydrometallurgical (wet) and pyroprocessing (dry) processes and assigned them a technology readiness level (TRL). Most of the processes are at level 4–6 of TRL, which means that technology was validated in a relevant environment (TRL 5) or was demonstrated in a relevant environment (TRL 6) but still is needed for the demonstration of the prototype in an operational environment.

3. Gen IV Reactors—Advanced Fuel Cycle—Perspectives

Certain parts of advanced fuel cycle concepts are still in the early stages of development (low TRL). This may involve greater uncertainties when compared to other options due to ongoing research. Reprocessing of spent nuclear fuel allows some of the components to be separated and used to produce new fuel that can later be used to generate energy. This might pose a challenge because of some new features of the advanced fuel that make it significantly different from conventional fuel.

3.1. High Temperature Reactors

High-temperature gas-cooled reactor (HTGR) is a helium-cooled graphite-moderated nuclear fission reactor technology considered a very safe, economical technology and environmentally friendly technology. The moderator and the coolant are extremely resistant to high temperature and high radiation activity. TRISO (TRi-structural ISOtropic) fuel used in the HTGR reactor is characterized by excellent fission product retention compared to the fuels from conventional reactors [24]. Moreover, TRISO fuel design promotes a very high burnup, significantly higher than the burn-up of current LWR fuels. This means TRISO fuel can be used for longer periods within a reactor before needing replacement, potentially leading to increased efficiency and reduced fuel cycle costs. In consequence, HTGR produces less HLW [25]. The TRISO fuel ensures inherent safety of the reactor, but it has a totally different structure from the fuels used today. These features make direct disposal of spent TRISO fuel currently the most convenient way to manage HLW from HTGR reactors. However, when considering radioactive waste from the HTGR reactor, not only the TRISO fuel can be considered, but also the irradiated graphite waste should be taken into account. TRISO particles are not used directly as fuel but are combined in fuel compacts. The individual HTGR fuel compact (spherical or cylindrical) consists of around a thousand TRISO particles embedded in a carbon matrix. Then the compacts are put into the graphite block [26]. The graphite matrix constitutes the major part of fuel elements as shown in Figure 4, therefore large volumes of irradiated graphite have to be considered at the back-end of the HTGR fuel cycle [27]. Moreover, the volume of the heavy metal contained in the HTGR fuel element is really small compared to the whole element. It is worth noting that the volume (not weight) is a key parameter in the radioactive waste management, together with the other ones as level of radioactivity, isotope composition, heat generation, etc.

Two paths can be adopted in the open HTGR fuel cycle: direct storage of whole graphite blocks or separation of spent TRISO particles from the graphite block, as shown in Figure 5 [28]. Direct storage of the full fuel block is currently the most commonly considered path for the management of HTGR spent fuel. The main disadvantage of this solution is the very large volume of high-level waste that requires enormous storage space where it can be safely disposed of for a long time. Spent fuel from HTGR reactors will have lower waste toxicity levels, compared to GENIII/III+ reactors, due to deep burning of plutonium and long-lived actinides in a single pass [29]. HTGRs are usually rated well; nonetheless, their sustainable development raises concerns due to the open fuel cycle. It could be addressed by the separation of the high-activity fuel compact from the low-activity graphite block [25,30]. The volume of high-level radioactive waste streams would be significantly reduced. Then, the management of spent fuel and the processing of graphite should be considered as separate pathways. Graphite could be classified as LLW [31]. It should be noted that the fuel blocks could also be recycled and reused. The compact with spent TRISO fuel could be further processed or intended for disposal in a geological repository [32]. The large blocks of spent HTGR fuel can be efficiently pulled out from the installation by using a mechanical method [33]. The TRISO particles are stable in this process. The electrolytic method was examined in the block and compact deconsolidation process but it does not remove pyrolytic carbon and silicon layers. Several studies have been carried out to optimize the electrolytic disintegration process but only in the lab scale [33,34,35,36,37,38,39,40,41,42]. Further volume reduction can be achieved by separating the HTGR fuel particles from the graphite matrix.

TRISO fuel particles are characterized by an extremely hard and highly refractory coating that prevents fission products from irradiated spent fuel. It is an advantage for future long-term storage but makes reprocessing options much more difficult. The methods used to reprocess spent fuel from LWR reactors cannot be easily adapted to reprocess spent fuel from HTGR reactors [43]. Various techniques have been investigated to separate particles from the matrix material and remove coatings (

Table 1. Technologies proposed for TRISO fuel processing.

	Utilization
Technology	Block-Compact Deconsolidation	Pyrolytic Carbon Removal	Silicon Carbide Removal	TRL [38]	References
Acoustical	No	Yes	Yes	1	[49]
Acid intercalation	Yes	No	No	2	[50]
Thermal shock	Yes	Yes	Yes	1	[51]
Pressure water jet	Yes	Partially removed	Partially removed	1	[52]
Hot chlorine gas	No	Yes	Yes	1	[33]
Pyrometallurgical method	No	Yes	Yes	1	[48]
Combustion	No	Yes	No	3	[33,40,41]
Electrolytic Constant current	Yes	No	No	2	[35,36,37,53]
Electrolytic pulsed current	Yes	Yes	Yes	2	[32,38,44,45,53,54]

). The first approach to reprocessing HTGR fuel was to mechanically crush the fuel particles and the combustion to remove the graphite [40,41,42]. The disadvantage of the process is the generation of huge amounts of off-gas, mainly carbon dioxide contains ¹⁴C [44]. Moreover, the SiC layer protects kernels from oxidation and it is not removed in this process. The release of the coated particles from the graphite matrix and the fragmentation process of coated particles were realized by using the pulsed current technology [33,45,46]. The fragmentation of TRISO particles was done with non-radioactive surrogates. It seems to be a promising technology, especially since no gas-off is generated, no chemical reagents are needed and it is not a high-energy-consuming process. The scale-up of the process is still necessary. Then, the continuation of tests with irradiated radioactive material should be carried out to demonstrate that the radiation does not affect the susceptibility of the graphite to fragmentation and the kernels remain intact in the separation process. However, the breach of these layers will result in the release of gaseous fission products. For this reason, additional operations will be necessary to capture, process, and dispose of these radioactive volatile products. All these operations should be included in a special front-end process aimed at removing coating layers. Nonetheless, after the fuel kernels are released from the silicon carbide and pyrolytic carbon layers, it will be possible to use the well-known actinide recovery technologies as PUREX or another aqueous process [11]. The integrating actinide recovery from spent TRISO fuel into the existing LWR fuel reprocessing installation seems to be a possible solution [47]. The pyroprocessing for actinides recovery from TRISO spent fuel is also investigated [48]. The separation of particles from the matrix material, followed by breaching the coatings, will be necessary to allow the molten salt electrolyte during pyroprocessing to penetrate and contact the uranium kernel. The other concept is the combination of hydro and pyro-processes for actinides recovery. This will be very difficult to implement from a plant perspective. Deconsolidation of graphite and breaching of particle coatings demands high temperatures and reagents other than those used for the recovery of actinides during the aqueous hydrometallurgy process. Pyroprocessing uses high-temperature molten salts, while aqueous reprocessing uses nitric acid to dissolve the spent fuel. In addition, the advantages of compactness and flexibility of the pyrochemistry process may be partially gone in the hybrid process.

Uranium monocarbide fuel is receiving growing interest as a preferred nuclear fuel composition for advanced reactors [55]. This fuel is characterized by higher conductivity and atomic density of the metal, which makes it easier to control during reactor operation. In this context it is worth noting that the recycling of actinides from uranium monocarbide fuel will be more complicated than the reprocessing of uranium oxide spent fuel. The uranium kernel is dissolved in a solution of nitric acid. Direct dissolution of the carbide fuel kernel results in the solution containing U and Pu contaminated by organic molecules that will interfere with metal complexation during liquid–liquid extraction. One of the proposed solutions to this issue was the peroxidation of uranium/plutonium carbide with CO₂ [51]. However, this process is potentially hazardous because the oxidation step is highly exothermic and involves the emission of ¹⁴CO₂. It seems that a transient mathematical model developed for UC pellet oxidation may be useful in this matter [56]. This simulation is a model that enables estimating the temperature, reaction rates, and the occurrence of any dangerous thermal course under diverse conditions. Additionally, the model allows for the analysis of different carbide processing routes, such as direct dissolution in nitric acid followed by removal of organic compounds from solution, pre-oxidation treatment to convert the carbide fuel to oxide, and then application of the Purex process.

3.2. Lead Cooled Fast Reactors

The crucial step in a fully closed cycle is the partitioning and transmutation (P&T). P&T can potentially recycle and reuse the separated minor actinides, contributing to a more sustainable nuclear fuel cycle. This process involves the partitioning (separating) radioactive materials from SNF and then transmuting them in nuclear reactions into much shorter-lived and stable nuclides that are less hazardous and more manageable forms. P&T is a complex operation involving either chemical or nuclear processes and requires advanced technologies and specialized facilities. This technology is not yet implemented but it is being widely researched and developed in several countries. Accelerator-driven systems (ADS) and fast neutron reactors (FNRs) are two technologies designed for transmutation [57].

A separate and complementary branch of advanced nuclear technology is represented by lead-cooled fast reactors (LFRs). In LFR systems, metallic fuels are used in conjunction with liquid lead pure lead (Pb) or lead–bismuth eutectic coolants (LBEs) [58]. Pure lead offers low neutron absorption, chemical inertness, and reduced activation products, particularly in minimizing radiotoxic isotopes post-irradiation. It forms a passivating PbO layer that protects structural materials from corrosion. In contrast, LBE has superior thermal conductivity and a low melting point (125 °C), enhancing operational flexibility. However, it generates polonium-210 (α-emitter, T_1/2 = 138.4 days) from neutron activation of bismuth-209, which poses significant challenges in radiation protection and coolant disposal. After decommissioning, solidified lead can be filtered and reused or disposed of as low-level waste. LBE, however, often requires immobilization techniques (e.g., vitrification or ceramic encapsulation) to safely contain polonium. Advanced methods such as controlled PoO₂ volatilization and cold-trap capture are under development [59,60]. Regardless of the coolant option you choose, LFR technology draws distinct advantages from the high-density, high-boiling-point, and excellent heat-transfer properties of lead, which contribute to superior safety margins and enhanced reactor performance under a fast neutron spectrum. Fast reactors operating in this manner exhibit a robust breeding capability, efficiently converting fertile isotopes into fissile material while simultaneously facilitating the transmutation of long-lived minor actinides [61]. Another advantage is that this reactor type can be engineered to operate in an adiabatic equilibrium state, a condition in which the isotopic vector of transuranic elements remains constant across refueling cycles [62]. In an adiabatic equilibrium cycle, the LFR converts natural or depleted uranium into plutonium and simultaneously consumes in situ-generated minor actinides. The spent fuel is reprocessed, and all actinides are recycled back into the core, leaving only fission products to be discharged. Three main conditions must be satisfied: (1) constant fuel composition cycle-to-cycle, (2) constant criticality at the beginning of each cycle, and (3) zero net breeding gain. These constraints allow for actinide mass stabilization, reducing long-term radiotoxicity. The European Lead-cooled Fast Reactor (ELFR), developed under the European Lead-cooled System (ELSY) and Lead-cooled European Advanced DEmonstration Reactor (LEADER) programs, demonstrated the feasibility of a two-batch reloading scheme, where fuel is irradiated for 1800 days and undergoes a 7.5-year cooling period before reprocessing [63,64]. Monte Carlo modeling using the MCB code showed that this strategy achieves equilibrium within a few cycles [62].

The fuel cycle associated with LFRs involves a series of processes that differ markedly from both conventional LWRs and molten salt reactors. Initially, metallic fuel elements are irradiated in a fast-neutron environment that promotes high burnup rates and effective breeding. Periodically, the irradiated fuel is reprocessed using pyrochemical techniques tailored for metallic fuels. Central to this reprocessing is an electrorefining stage conducted in molten salt media, where actinides are selectively recovered from the spent fuel. Solid aluminum cathodes, often enhanced by porous structures, are typically employed to deposit stable actinide–aluminum alloys. This selective deposition is supported by careful control of operating potentials and by advanced real-time monitoring systems that ensure process stability despite the presence of residual lead and other contaminants [65].

A critical distinction of the LFR fuel cycle is its emphasis on closing the fuel cycle by recycling nearly all transuranic material. By recovering nearly the entire inventory of actinides and returning them to the reactor, LFRs can reduce the overall mass of waste significantly, by factors of up to 20 compared to LWRs. Furthermore, combined with the use of advanced uranium nitride (UN) fuel and lead-based coolants, this concept supports a fully closed fuel cycle with minimal external inputs and waste outputs. UN fuel is increasingly favored for LFRs due to its high uranium density (~13.55 g/cm³), superior thermal conductivity (~20.6 W/m·K), and compatibility with fast spectra [66]. These features enable high breeding ratios and compact core designs, essential for achieving adiabatic conditions [67,68]. However, the chemical reactivity of UN makes aqueous reprocessing (e.g., PUREX) infeasible. Pyroprocessing in molten salt media (typically LiCl-KCl) offers a viable alternative [69]. Spent UN fuel is electrochemically dissolved, and actinides are selectively recovered. This dry process is thermally compatible with nitrides and avoids complications from water-sensitive matrices [70]. An additional challenge is the recovery of enriched nitrogen-15 to suppress carbon-14 production during irradiation. This necessitates integration of ¹⁵N recovery systems into the head-end processing steps [71]. The fast spectrum inherent to LFRs means that the residual radiotoxicity in waste forms declines much faster, potentially reducing the required isolation period from tens of thousands of years to only a few centuries. The LFR cycle also offers enhanced proliferation resistance by diluting weapon-usable material within a broader, chemically stable matrix. Ongoing research emphasizes the need for further improvements in electrode design, material durability, and dynamic process control to scale these processes efficiently [72,73].

The integration of LFR technology within an advanced nuclear energy strategy is not intended to replace LWRs altogether; rather, it is envisioned as a complementary component that can process the spent fuel from traditional reactors or operate as part of a diversified reactor fleet. An optimal nuclear energy strategy may involve the gradual transition from a predominantly once-through cycle to a more closed advanced cycle that leverages both MSR and LFR technologies. Such diversification would allow for more efficient resource use, expansion of breeding and transmutation capabilities, and a marked reduction in the volume and longevity of high-level radioactive waste.

3.3. Molten Salt Reactors

Molten salt reactor (MSR) family combines the inherent advantages of liquid-fuel technology with a strategy for “destroying” long-lived transuranic actinides. In a conventional MSR—as first demonstrated by molten-salt reactor experiment (MSRE) at Oak Ridge National Laboratory in 1965–1969, a fuel and coolant were dissolved together in a circulating fluoride–salt mixture (e.g., LiF–BeF₂–UF₄) that flows through a graphite-moderated core at atmospheric pressure and temperatures approaching 600–700 °C. Building on this concept, the molten-salt transmutation reactor (MSTR) design was elaborated [74]. MSTR represents a specialized evolution within the broader class of molten salt reactors, designed explicitly to address the long-standing challenge of managing long-lived transuranic elements in spent nuclear fuel. MSTR utilizes a liquid fuel form—typically a eutectic mixture of fluoride salts containing dissolved actinides, which enables continuous reprocessing and dynamic fuel management. In the MSTR paradigm, front-end reprocessing of spent LWR oxide fuels via the fluoride volatility method, leaving non-volatile lanthanide and transplutonium salts behind, thereby concentrating actinides for subsequent dissolution into MSR salt [74,75]. Within the reactor’s primary loop, continuous salt-to-metal reduction methods (e.g., molten-salt/liquid-bismuth extraction) remove large neutron-absorbing lanthanides, while electrochemical separation techniques recover any remaining fission products and redistribute actinides back into the core [76,77]. This unique configuration allows for the in situ transmutation of minor actinides such as neptunium, americium, and curium, converting them into shorter-lived or stable isotopes, thereby significantly reducing the long-term radiotoxicity of nuclear waste [74]. The reactor operates in a thermal or epithermal neutron spectrum and is designed to accommodate a high amount of transuranic isotopes, which are continuously circulated through the core and subjected to neutron irradiation. The liquid nature of the fuel permits the extraction of fission products and the reinsertion of unburned actinides without interrupting reactor operation. This continuous fuel processing loop is central to the MSTR’s ability to maintain a favorable neutron economy while minimizing the accumulation of neutron poisons [78,79]. This configuration of the technology also permits continuous “on-line” removal of gaseous and solid fission products, thereby maintaining reactivity and extending fuel life. Although MSRs offer a promising alternative to conventional nuclear designs, the technology raises several critical issues that must be addressed for the technology to mature and reach commercial viability.

One of the foremost challenges is the interaction between the molten salts and structural materials. Operating at high temperatures, MSRs expose reactor components to an aggressive environment that can lead to accelerated corrosion and material degradation. The chemical compatibility of molten salts with metals and alloys is complex, requiring advanced materials research to develop solutions such as novel alloys or protective coatings that can endure long-term exposure to both high temperatures and intense radiation. This material challenge is essential not only for maintaining structural integrity but also for ensuring the overall safety of the reactor system [80,81].

Thermal management represents another critical concern. MSRs are designed to operate at temperatures significantly higher than traditional water-cooled reactors, which can substantially improve thermal efficiency. However, the high-temperature operation demands meticulous control of heat transfer processes to avoid localized overheating and thermal stresses. Additionally, the behavior of the circulating fuel salt under dynamic conditions must be accurately predicted to ensure stable operation during both routine functioning and unexpected transients. Successfully addressing these thermal-hydraulic challenges is essential to harnessing the full performance potential of MSRs while maintaining safety margins [81,82].

One of the primary technical challenges in the reprocessing of spent fuel from MSRs is the chemical complexity of the molten salt medium. The fuel salt typically contains a mixture of actinide and lanthanide chlorides or fluorides along with a host of fission products. The process of separating these components requires highly selective and robust chemical processes. Unlike solid fuel reprocessing, where physical separations (e.g., PUREX process) are well-established, the chemical separation in a molten salt medium faces additional hurdles due to the high temperature and corrosivity of the salt mixture. The research focuses on developing solvent systems and electrochemical processes that can perform with high selectivity and chemical stability in molten fluoride or chloride salt matrices, typically within the 500–800 °C temperature range.

Basically, two pyrochemical routes have been defended for MSR fuel management. First, in the MSTR model, front-end reprocessing of spent LWR oxide fuels via the fluoride volatility method converts UO₂ and mixed oxides into UF₆ while leaving non-volatile lanthanide and transplutonium salts behind, thereby concentrating actinides for subsequent dissolution into MSR salt [74,75]. Within the reactor’s primary loop, continuous salt-to-metal reduction methods (e.g., molten-salt/liquid-bismuth extraction) remove large neutron-absorbing lanthanides, while electrochemical separation techniques recover any remaining fission products and redistribute actinides back into the core (Figure 6). Second, electrochemical separation in molten fluoride media—most often LiF–NaF–KF (FLINAK) or LiF–CaF₂—utilizes finely tuned redox potentials to plate out actinides on specialized cathodes (e.g., solid Al or W electrodes) while leaving fission products in solution [74,76,77,83]. These techniques have been demonstrated in laboratory-scale cells—reaching technology readiness levels of approximately 3–4, but still require prototypic loop testing under irradiation to achieve TRL 6 and above [56].

One promising path involves the use of ionic liquids, which are room-temperature molten salts known for their thermal stability, small vapor pressure, and capacity to solvate metal ions. When adapted to high-temperature conditions, these materials enable the development of extraction systems that can selectively partition actinides from a complex mixture of fission products. The design of ionic liquid-based extractants often incorporates customized chelating ligands such as bis-triazinylpyridines, phosphine oxides, and functionalized diamides that exhibit strong selectivity for trivalent actinides over lanthanides. The challenge lies in ensuring that these ligands remain chemically stable in aggressive molten salt environments without significant degradation or radiolysis [43,78]. In parallel, electrochemical methods, particularly electrorefining and electroreduction, are being optimized for molten salt applications. These processes use the redox potential differences between actinides and fission products to achieve selective deposition. For instance, in a LiCl–KCl eutectic mixture, uranium and transuranic elements can be reduced at cathodes, while lanthanides and other fission products remain in solution. The control of redox potential range is critical, and current research focuses on advanced electrode design, including the use of porous cathode structures and dynamic voltage modulation to enhance deposition efficiency and minimize co-deposition of impurities [53,72,76]. Moreover, researchers are exploring hybrid systems that merge electrochemical and solvent-based separation strategies. These may involve initial electrochemical pre-separation followed by solvent extraction refinements, enabling multistage purification and minimizing the generation of secondary waste. In situ monitoring tools, such as cyclic voltammetry and spectro electrochemical sensors, are increasingly integrated into these setups, offering real-time feedback and adaptive process control [84].

Another issue is handling fission products and waste streams. In advanced molten salt reactor systems, the continuous or semi-continuous reprocessing of fuel salt generates a variety of fission product waste streams that differ significantly from those produced in conventional solid-fueled reactors. These waste streams include volatile species (e.g., iodine, cesium, and noble gases), semi-volatile metals (e.g., tellurium, selenium), and refractory elements (e.g., molybdenum, zirconium, rare earths). The removal of fission products is a key requirement, and their efficient separation is imperative for maintaining reactor criticality. Many fission products exhibit significant neutron absorption cross-sections, and if left unchecked, they can act as strong poisons. The chemical separation process must, therefore, extract these isotopes with high purity to ensure the fuel salt remains conducive to sustaining a controlled nuclear reaction. The complexity increases when multiple fission products require simultaneous separation. The traditional PUREX process is not directly applicable to a high-temperature salt medium. Novel separation techniques such as advanced solvent extraction methods tailored for fluoride and chloride melts are being developed. These processes must achieve a delicate balance: they should remove neutron poisons efficiently while also minimizing the production of secondary waste streams that require their own management and disposal strategies. The issue of secondary waste is further compounded by the chemical and radiological properties of the separated fission products, which might necessitate long-term storage in stable waste forms [85]. Stabilizing into durable waste forms is essential for ensuring the long-term safety and environmental acceptability of MSR technology. Current research focuses on the immobilization of the fission products into chemically stable, leach-resistant matrices such as phosphate-based glasses, glass–ceramics, and crystalline ceramics [86]. In addition, by embedding transmutation directly within the reactor core, the MSTR offers a pathway to drastically reducing the repository burden of minor actinides and harnessing energy from legacy LWR fuel stocks. Comparative neutronics studies show that, on an equilibrium basis, fast-spectrum MSTR designs can achieve minor-actinide transmutation rates comparable to sodium or lead fast reactors, with the added advantage of in-core online reprocessing and a lower inventory of unburned actinides [77,78,87]. It is assumed that such reactors can achieve transmutation rates exceeding 250 kg/year of minor actinides at thermal power outputs of 2400 MW. These findings underscore the reactor’s potential to serve as a dedicated burner for long-lived isotopes extracted from spent LWR fuel. Moreover, the modular loop configuration proposed for MSTRs enhances thermal management and minimizes the volume of fuel salt outside the core, thereby improving both safety and transmutation efficiency [79].

Integrating the reprocessing system with reactor operation is inherently complex. In MSRs, continuous reprocessing is favored to avoid large-scale disruptions in reactor criticality. However, establishing a continuous loop for salt withdrawal, chemical treatment, and salt return without inducing operational instability poses significant process control challenges. The overall system must accommodate fluctuations in salt composition, temperature, and beam radiation effects without compromising safety or efficiency. Real-time monitoring and advanced sensor technologies could play an important role in maintaining process control. However, the harsh operating conditions hamper the reliability and longevity of standard sensors. Developing resilient sensor systems that can provide real-time data on chemistry, temperature, and radiation in molten salt environments is an area of active research. These systems are essential for ensuring consistent reprocessing rates and preventing reactor shutdowns due to process variability.

Material compatibility and corrosion issues at the moment are one of the key challenges in MSR technology. The development of corrosion-resistant materials is a keystone of enabling long-term, safe, and economically viable operation of molten salt reactors. The extreme chemical reactivity of molten fluoride and chloride salts at elevated temperatures (typically 500–800 °C), combined with intense neutron irradiation, presents a uniquely aggressive environment for structural materials. MSRs require materials that can resist both mechanical wear and chemical corrosion. The molten salt medium is inherently aggressive, attacking conventional steels and alloys. Reprocessing equipment, including vessels, piping, and separation units, must be constructed of materials that can withstand prolonged exposure to corrosive salts and high radiation doses [88]. Corrosion not only shortens the lifespan of reactor components but also introduces impurities into the fuel salt. Innovative materials such as Hastelloy-N have been utilized in experimental settings; yet, scaling these materials up for commercial deployment remains challenging. The development of corrosion-resistant coatings and advanced alloys is a crucial step towards ensuring the reliability of processes used for salt reprocessing [89]. Recent studies have also explored the addition of elements such as niobium, silicon, and manganese to further enhance oxidation resistance and reduce mass loss under prolonged exposure to molten salts. In parallel, ceramic and refractory coatings are being developed to serve as diffusion barriers and sacrificial layers. These include yttria-stabilized zirconia, silicon carbide, and alumina-based coatings, which can be applied via chemical vapor deposition or plasma spraying. These coatings aim to isolate the base metal from direct contact with the salt, thereby reducing corrosion rates and extending component lifetimes. Some coatings are also being engineered with self-healing properties, where embedded reactive phases can re-passivate damaged regions by forming protective oxide layers in situ [90]. Moreover, intensive research on real-time monitoring systems to track corrosion levels and salt composition can provide early warning signals for system maintenance and prevent unexpected failures.

A consolidated TRL survey [56] indicates that MSR core technologies—graphite moderation, salt handling, and basic reactor physics—are at TRL 4–5 (bench-scale validation in MSRE). FVM for uranium extraction has reached mid-TRL 5 with pilot-scale fluoride-gas reactors, whereas salt-to-metal extraction and electrochemical purification are at TRL 3–4 following laboratory demonstrations [74,76]. To advance the MSTR system to TRL 7–8—sufficient for a lead test assembly or demonstration reactor—integrated loop testing under prototypical neutron flux, temperature, and flow conditions will be required, alongside corrosion qualification of structural alloys to TRL 6–7 and sensor validation under irradiation [91,92].

3.4. Accelerator-Driven System

ADS are nuclear reactors that utilize a particle accelerator to produce neutrons, which then induce fission. ADS is being explored to transmute minor actinides into shorter-lived or stable isotopes [93,94]. It can coexist with fast reactors or play a complementary role, as shown in Figure 7.

In the first approach (Figure 7a,b) the ADS is the tool to transmute minor actinides produced in conventional reactors (like LWRs). The ADS has the advantage compared to FNR that it can efficiently burn pure minor actinides while maintaining the core safety characteristics. Fast reactors are also capable of burning actinides, but there may be design and fuel safety issues with pure minor actinide fuels. The high neutron absorption cross-sections and reactivity of minor actinides can lead to increased peaking power factors, potentially affecting core stability and safety.

The strategy called double strata involves using different types of reactors or systems in a complementary manner. There are two layers. The first strata involves the transmutation of plutonium and other actinides, usually in a fast reactor, which results in energy production and the reduction of the amount of long-lived radioactive material.

Then, minor actinides that may not have been fully transmuted in FNR, e.g., americium, curium, can be subjected to further transmutation in ADS (Figure 7c). The double-strata strategy is a layered approach to nuclear waste management that merges the advantages of various reactor technologies and allows for achievement of more efficient and effective transmutation of radioactive materials. Reducing the radiotoxicity of HLW can reach an even factor of one hundred compared to the open fuel cycle [5].

3.5. Thorium Cycle

Thorium is significantly more abundant than uranium in the Earth’s crust and it is a promising nuclear energy source. The thorium fuel cycle has been extensively studied and demonstrated in research reactors. It offers several advantages over the traditional uranium-plutonium cycle, among them potentially lower proliferation risk, higher fuel efficiency, and reduced production of long-lived transuranic elements.

The thorium fuel cycle presents a compelling alternative to the conventional uranium–plutonium cycle, offering advantages in resource abundance, waste minimization, and proliferation resistance. However, its implementation across different reactor technologies requires tailored reprocessing strategies and careful consideration of fuel behavior, isotopic evolution, and material handling challenges.

Thorium-232 is a fertile isotope that, upon neutron absorption, transmutes into fissile uranium-233 via the intermediate protactinium-233. Unlike uranium, thorium contains no naturally occurring fissile isotopes, necessitating an external neutron source or fissile isotopes (e.g., ²³⁵U or ²³⁹Pu) to initiate the cycle. Once established, the ²³²Th–²³³U cycle can be closed through reprocessing and recycling of ²³³U, enabling thermal breeding in suitable reactor configurations [95,96]. Reprocessing thorium fuel involves unique challenges due to the chemical inertness of ThO₂, the radiotoxicity of ²³²U decay products, and the need for remote handling. Two primary reprocessing routes have been explored.

(1)

Aqueous reprocessing (THOREX process) adapted from the PUREX process; THOREX uses tributyl phosphate in nitric acid to extract ²³³U from irradiated ThO₂. However, ThO₂’s low solubility in HNO₃ complicates dissolution, often requiring fluoride additives or high-temperature treatment. The presence of ²³²U and its gamma-emitting daughters necessitates shielded facilities and remote operations [76].

(2)

Pyrochemical reprocessing for high-temperature reactor systems or molten salt fuels; pyroprocessing in molten fluoride or chloride salts offers a viable alternative. Techniques include electrorefining, fluoride volatility, and reductive extraction, as was mentioned in previous chapters [74,97].

The thorium fuel cycle can be used in different reactor technologies. In light-water reactors, particularly pressurized water reactors (PWRs) and boiling water reactors (BWRs), thorium has been incorporated through seed-and-blanket designs, where a fissile material such as ²³⁵U or ²³⁹Pu ignites the chain reaction and thorium acts as a fertile blanket material. Though technical feasibility has been demonstrated—including in historical trials such as the Shippingport LWBR—the inability to recycle ²³³U and the presence of strong gamma-emitting ²³²U isotopes have impeded widespread commercial deployment. In these systems, reprocessing would necessitate adaptation of aqueous chemical flowsheets like THOREX, a variant of PUREX specifically configured to separate uranium from thorium dioxide fuel matrices. However, the low solubility of ThO₂ in nitric acid and the radiological hazards posed by ²³²U daughters demand rigorous shielding and remote operation facilities, thereby complicating economic viability [98,99].

High-temperature gas-cooled reactors, including both pebble-bed and prismatic-core designs, offer another platform for thorium utilization. These reactors often deploy TRISO (tri-structural isotropic) coated particle fuels containing ThO₂ kernels. The resilience of TRISO particles under extreme temperatures makes them suitable for thorium incorporation, but it also renders reprocessing highly challenging. Chemical extraction of ²³³U from TRISO fuels involves intricate disassembly and treatment of multi-layered ceramic coatings. As a result, the fuel cycle in HTGRs with thorium is typically designed as a once-through system, though research continues into hybrid mechanical–chemical reprocessing paths [100].

In molten salt reactors (MSRs), thorium exhibits unique synergies with the liquid fuel format. Thermal spectrum MSRs, particularly those employing a thorium–uranium fuel dissolved in fluoride-based salts, allow for real-time adjustment of fuel composition and the removal of fission products via sparging, volatility, or batch chemical processes. The Molten Salt Breeder Reactor concept demonstrated potential for breeding ²³³U from thorium with continuous online reprocessing. The process involves isolating volatile species, removing neutron poisons, and separating actinides and lanthanides via reductive extraction or electrodeposition. However, challenges related to corrosive salt chemistry, sensor integration, and high-temperature materials durability have kept the TRL of a complete MSR thorium cycle at approximately 4–5 [77,83].

Fast reactors, such as lead-cooled (LFR) or gas-cooled (GFR) variants, have considered thorium as both a blanket material and fuel matrix in mixed oxide configurations. These designs aim to transmute surplus plutonium while simultaneously breeding uranium-233. Although experimental assemblies using (Th,Pu)O₂ fuels have been irradiated—particularly in Russian SFR programs—the reprocessing technologies lag behind in maturity. Pyroprocessing in molten chloride salts is preferred due to its compatibility with metallic and refractory oxide fuels, and its ability to operate under high radiation fields. Electrochemical partitioning of thorium, however, requires precise redox control and remains at the bench-scale demonstration stage [56,76]. Accelerator-driven systems (ADSs) also utilize thorium’s favourable transmutation properties. In these subcritical reactors, external spallation neutron sources permit high-fluence irradiation of thorium targets, effectively converting ²³²Th into ²³³U without the need for maintaining core criticality. Though still conceptual, ADS designs incorporating thorium envision closed fuel cycles supported by pyrochemical reprocessing under high radiation and thermal loads. The TRL of such systems remains low, though their potential for reducing actinide inventories is acknowledged in strategic fuel cycle studies [96].

In summary, while the thorium fuel cycle offers substantial long-term benefits, including lower actinide production, enhanced safety margins, and proliferation resistance, it poses significant technological barriers that vary by reactor type. The thorium spent fuel requires a longer cooling time than uranium spent fuel because ²³³Pa is produced in the conversion chain of ²³²Th to ²³³U. The presence of protactinium in waste from the processing of spent thorium fuel may also have long-term radiological impact. Reprocessing of thorium-based fuels remains complex due to the inertness of ThO₂, the gamma-emitting decay chain of ²³²U, and the need for advanced chemical separations and remote handling infrastructure. Across reactor systems, TRLs for thorium fuel cycles generally range from 3 to 5, with higher readiness observed in MSR subcomponents. Continued investment in reprocessing science, sensor integration, and materials engineering will be essential to elevate these systems toward commercial deployment. The thorium cycle still faces challenges in implementation and commercialization. Intensive research and development across the entire fuel cycle is required before significant investment can be made in the commercial use of thorium fuels and fuel cycles [95].

4. Discussion

Nuclear power is considered a significant energy source for the future, as it can help meet growing electricity demands at the same time simultaneously reducing carbon dioxide emissions. However, the most popular fuel cycle in most of the world today may be considered unsustainable. Disposal of SNF generated by commercial nuclear power plants operating in an open fuel cycle in a deep geologic repository is the only viable solution. However, there is still no operational deep repository at present. The Waste Isolation Plant in the USA is the only one licensed deep geological repository in the world but it is not licensed to dispose of HLW. The plans are also very advanced in other countries such as Finland, Sweden, and France [101]. SNF after disposal of the reactor is kept at the reactor pool for about 5–10 years to cool and then moved to dry casks and transported to interim storage, which is a dedicated storage for another 40–50 years to cool enough to dispose of underground. All generated so far, HLW remains in interim storage waiting for an operating geological repository. It should be noted that space is limited. The situation may become even more serious as many countries invest in advanced small modular reactors. Topic of the back-end management of the fuel that comes out of them should be considered in parallel with construction aspects. Minimizing the volume of high-level waste is key to the sustainable development of nuclear energy. The closing of the fuel cycle is the way to achieve this goal [102].

The transmutation using a single reactor concept may not be sufficient to burn all actinides effectively. The study, conducted as part of the EU PUMA project [101,103], examined the potential of closing the nuclear fuel cycle by utilizing SNF from LWRs in High HTRs and GCFRs. In general, symbiotic fuel cycles involve creating a closed system in which the spent fuel from one reactor type serves as serves as fuel to another, maximizing resource utilization and minimizing waste LWRs are currently the most popular type of nuclear reactor and appear to be the best choice for the initial reactor in first symbiotic nuclear cycles (Figure 8).

The MSR technology also offers a promising pathway for achieving enhanced fuel utilization and improved safety compared to traditional nuclear reactor designs. Under the advanced fuel cycle model, MSRs are proposed to close the fuel cycle by continuously reprocessing spent fuel, potentially reducing the volume and toxicity of nuclear waste. The advanced fuel cycle in the context of MSRs suggests that spent fuel can be continuously processed in situ to remove fission products and to recover fissile and fertile materials. This continuous reprocessing is fundamentally different from the batch-mode reprocessing required by traditional solid-fuel reactors and promises to reduce the amount of long-lived actinides and critical waste components. The advanced fuel cycle does not simply extend the life of the fuel; it promises to dramatically reduce the volume and radiotoxicity of high-level waste by converting long-lived actinides into shorter-lived fission products. For molten salt reactors, the continuous reprocessing approach theoretically enables a “closed-cycle” system. This approach may contribute not only to improved fuel utilization but also to enhanced proliferation resistance since the chemical forms in the salt differ significantly from those in conventional solid-fuel systems. The interest under the Generation IV International Forum has focused R&D on both “thermal” ²³²Th→²³³U cycles and “fast” transuranic-burning cycles [74,77]. However, the chemical complexity of molten salt fuel, combined with the harsh operating conditions, radiation fields, and corrosive environments, presents significant challenges to the reprocessing of spent fuel streams.

The significant benefits of recycling in waste management and resource use also come with some challenges [104]. The reprocessing of spent fuel in any nuclear system carries inherently radiological and proliferation risks. In continuous reprocessing schemes, material accounting and safeguards must be of the highest standard. Although MSRs were initially heralded for their potential proliferation resistance—owing to the difficulty of extracting weapons-usable material from a complex salt matrix—the practical realities of chemical separation processes may open vulnerabilities if appropriate controls are not implemented. Advanced process monitoring and automation, combined with robust nondestructive assay techniques, are critical to ensuring that reprocessed material is managed safely and securely.

Moreover, the use of new types of fuel such as thorium, UN, TRISO, and metallic fuels brings both many opportunities and many technological challenges. Reprocessing these types of fuels still requires research and testing. The decision of whether to close the fuel cycle through reprocessing or pursue other options is complex and depends on various factors, including security concerns, technological advancements, and national energy policies.

5. Conclusions

Reprocessing and reusing spent nuclear fuel has been a strategy used by some countries for many years. Despite its high costs and the need for advanced infrastructure and expertise, it is a promising path to achieving a more sustainable and efficient energy source. It maximizes the energy potential of nuclear fuel, more efficiently utilizes uranium resources, and potentially reduces the need for mining. Furthermore, this strategy helps minimize the amount of waste requiring long-term storage. With the development of advanced reactor systems, such as fast reactors, reprocessing could play an even more important role in the future of nuclear energy.

From a systems perspective, the HTR/FNR or MSTR is proposed as a key component in a hybrid nuclear infrastructure, complementing conventional reactors by consuming their most problematic waste streams. The integration of advanced reprocessing into a closed fuel cycle not only reduces the need for geological repositories but also enables the recycling of valuable actinides, contributing to resource sustainability. Nevertheless, several challenges remain. While significant progress has been made in understanding the challenges associated with advanced fuel reprocessing, many questions remain. Pilot-scale experiments and integrated test facilities are essential for verifying laboratory research under realistic operating conditions. The development of robust computational models that simulate the complex chemical interactions in SNF and advanced coolants will be instrumental in designing next-generation reprocessing systems.

Advancing to high TRL levels will require integrated loop demonstrations at engineering scale, material qualification under prototypic irradiation and flow conditions, and regulatory framework development for handling liquid fuels and salt wastes.

Ongoing research into advanced separation techniques, the development of corrosion-resistant materials, and the integration of real-time monitoring systems represents a concerted effort to address these challenges. With a robust collaborative framework and innovative technological approaches, the obstacles to reprocessing spent fuel may gradually be surmounted, paving the way for a cleaner and more sustainable nuclear future.