Today, thermal reactors, which are the basis of modern Russian nuclear energy, can consume plutonium, but in limited quantities and with a low content of threshold isotopes, so their fuel supply is actually limited by natural uranium reserves. Fast reactor technology was created for extended production of plutonium and their fuel supply capabilities are limited by the resources of 238U, which is two orders of magnitude more than 235U. The paper presents the following possible functions of fast reactors in the future two-component nuclear energy system, the transition to which in the coming decades is a strategic goal of the Rosatom State Corporation: electricity generation; maintenance the balance of plutonium in the system of fast and thermal reactors; “refinement” of plutonium from spent nuclear fuel in thermal reactors; disposal of long-lived nuclear energy wastes; production of isotopes highly demanded by the industry and medicine. Obviously, the specialized fast reactors creation will be required to increase the efficiency of these functions. As a result, thermal and fast reactors can be used to create a sustainable, multi-functional and waste-free new nuclear power industry with an almost unlimited fuel resource.

Two-component nuclear energy system, fast reactors, thermal reactors, spent nuclear fuel, closed nuclear fuel cycle, plutonium, uranium, minor actinides, isotopes utilization

The strategic goal of the State Atomic Energy Corporation ROSATOM for the coming decades is the transition of the nuclear power industry to a two-component structure with thermal and fast reactors and a single closed nuclear fuel cycle (NFC) (Alekseev et al. 2012; Alekseev et al. 2016; Alekseev et al. 2018; Alekseev et al. 2019b; Adamov 2020; Zrodnikov et al. 2020; Alekseev et al. 2022; Strategy 2021 2023). The two-component nuclear energy system (NES) is expected to address the following systemic issues (Troyanov et al. 2024):

• to provide long-term and reliable nuclear power fuel supply via conversion from uranium fuel to mixed uranium-plutonium fuel;
• to manage and balance the plutonium reserves in the global nuclear energy system with thermal and fast reactors, including “refinement” of the plutonium isotope vector;
• to reprocess spent nuclear fuel (SNF) from thermal reactors, including SNF repatriated from foreign Russian-designed nuclear power plants;
• to utilize minor actinides and other long-lived high-level waste accumulated as a result of thermal reactor operation;
• to produce medical, industrial and energy-generating isotopes, including cobalt-60, plutonium-238, and others.

Depending on their functional tasks (IAEA 2012; Lemasson et al. 2012; Alekseev et al. 2019a; Risovanyi 2020; Kotov et al. 2023; Kashirina et al. 2025), fast reactors can be divided into a number of groups, as proposed in (Camarcat et al. 2011) for fast reactors designed for electricity generation and for minor actinide (MA) burner reactors. In addition to these types of reactors, the Russian nuclear energy system in the future may also include specialized reactors for fast and thermal reactors’ fuel breeding, “conditioner” reactors for “refinement” of the isotopic composition of plutonium from thermal reactors, “isotope production” reactors to produce various isotopes both in the core and blankets and, possibly, others (Troyanov et al. 2024). The designs of the above reactors may vary in view of the specific tasks to be solved and the corresponding technical requirements for the fuel, core, and fuel infrastructure. All together they constitute a multifunctional component of fast reactors in the two-component nuclear energy system.

Currently, only series-built high-power sodium fast reactors (Poplavsky et al. 2010; Alekseev et al. 2018) can be considered as the existing technological basis for such a component, which along with production of electricity are capable of closing a single fuel cycle with thermal reactors and solving the above problems.

Let us consider the functions of sodium fast reactors in more detail.

The main function of fast reactors is the production of electricity at its lowest possible levelized cost (LCOE) (rated/averaged cost of electricity (from English Levelized Cost of Energy)). Fast reactors use a liquid metal coolant, which allows for higher steam temperatures (compared to VVER reactors). Therefore, these reactors produce electricity with higher efficiency, i.e. with lower specific consumption of heavy metal (uranium) and fission product generation.

The LCOE reduction can be achieved by reducing the capital costs, shortening the power unit construction time, as well as due to minimizing the operational expenses, cutting down the fuel cost component (FCC) as a result of optimization of BN-reactor SNF reprocessing, and reducing the MA handling, including organizational and economic measures.

As an example, Ref. (Tuzov et al. 2022) shows that the FCC reduction can be achieved by making a decision on “deferred” BN spent fuel reprocessing until there is a shortage of plutonium in the course of the intensive startup of fast reactors, and the reprocessing is more economically feasible. It is assumed that in this case the entire amount of plutonium from the SNF of thermal reactors will be recycled once through sodium fast reactors, and the accumulated amount of SNF of BN-type reactors will be sufficient for the economically justified construction and capacity utilization of the centralized reprocessing plant.

This issue was studied for a model scenario of nuclear energy development in Russia (Tuzov et al. 2022). The scenario is shown in Fig. 1a. The results according to Fig. 1b demonstrated that the need for reprocessing SNF of fast reactors in the scenario under consideration will arise after 2055. At the same time, the FCC in electricity production can be reduced by 25% as a result of prioritized and large-scale reprocessing of SNF of thermal reactors.

Figure 1. 

The structure of SNF reprocessing capacities (a) and volumes (b) in the two-component NES development scenario with thermal and fast reactors (Tuzov et al. 2022).

The next function is related to the feasibility of plutonium management and balancing in the two-component NES. It is based on a fast reactor ability to both burn and breed plutonium.

Table 1 schematically shows the considered options for the blankets in a model high-power sodium fast reactor.

In each option, reprocessing of the core SNF is carried out with consideration of plutonium from all the blankets. Therefore, for each option there is its own equilibrium fuel.

Fig. 2a–e shows the breeding ratios, balance of plutonium, fraction of its odd isotopes, mass of plutonium in the course of reprocessing in a closed nuclear fuel cycle (CNFC) of the reactor, production of americium and curium for all the options, respectively. It can be seen that as the number of blankets increases, the breeding ratio changes from 0.98 to 1.12, and the importance of the fuel improves– the fraction of odd isotopes increases from 0.61 to 0.72, and the excess plutonium after BN SNF reprocessing exceeds 100 kg/year.

Figure 2. 

Breeding ratios (a), plutonium balance (b), fraction of odd plutonium isotopes (c), reprocessing plutonium mass (d), production rate of americium (e) and curium (f) in the reactor fuel cycle for options 1 ÷ 7, respectively.

The effect of maximum involvement of plutonium from the blankets into the fuel cycle can be described as fast reactor plutonium “self-refinement”, which makes it possible to minimize the amount of plutonium and its recycling in the closed nuclear fuel cycle of the reactor by reducing the fraction of threshold isotopes from 39% to 28%. In addition, plutonium fission occurs in the radial blanket of the reactor in the course of plutonium production, the thermal power released in this case is ~ 3% of the total reactor power, which makes it possible to relieve the fuel assemblies of the core and increase its fuel residence time. This may also have a positive effect on LCOE FCC.

In addition, as the breeding properties of the “clean” fuel improve, the production rate of americium decreases according to Fig. 2e and that of curium decreases according to Fig. 2f.

Another positive effect of the maximum application of blankets consists in the fact that the excess plutonium from the fast reactor radial blanket can be used without restrictions on the number of threshold isotopes as MOX-fuel for future thermal reactors with spectral regulation, such as VVER-S or VVER-SKD. Since the fuel enrichment (in plutonium) in a thermal reactor is ~4–5 times lower than in a fast reactor, the amount of such fuel will be quite significant. Thus, a radial blanket of a BN-type reactor can produce up to 100 kg of low-burnup plutonium per year, which is sufficient to fabricate ~2 tons of thermal reactor fuel.

Plutonium from thermal reactor SNF is characterized by a high content of threshold isotopes and cannot be reused in them. In addition, the breeding ratio of thermal reactors is less than unity. Therefore, a closed fuel cycle of uranium-plutonium fuel in thermal reactors is impossible. At the same time, when plutonium from the thermal reactor SNF is used in fast reactor fuel, the content of threshold isotopes is significantly reduced. In addition, there are several blankets in a fast reactor: the axial fertile plate, bottom and radial blankets where low-burnup plutonium is produced from depleted uranium. The idea to adjust the isotopic composition of plutonium (from PWR SNF) through its irradiation in fast reactors (using the example of ASTRID) was proposed by the French experts in 2017 (Grouiller et al. 2017; Camarcat et al. 2019). When this plutonium is mixed with plutonium from irradiated fast reactor fuel, the fraction of threshold isotopes is further reduced, which allows this plutonium to be used in thermal reactors. The mass of this “refined” plutonium will be an order of magnitude greater than that of plutonium from the radial blanket, which will allow a joint nuclear fuel cycle to be arranged from thermal and fast reactors. The R&D conducted in Russia have confirmed the efficiency of such a joint fuel cycle (Eliseev et al. 2020; Klinov et al. 2020). The studies have shown that one fast reactor will provide fuel for 1 MOX-fueled VVER-S-600 reactor, or 2 VVER-1200 reactors operating on 1/3 MOX-fuel. In addition, it is expected that there will be a possibility of rendering services on refinement of plutonium from the SNF of PWR with MOX-fuel of a foreign customer. The ratio of VVER and BN-type reactors in a joint CNFC is shown in Fig. 3.

Figure 3. 

The ratio of VVER and BN-type reactors in a joint CNFC.

The results of joint fuel cycle modeling show that the refinement of plutonium in the MOX-fuel of a high-power sodium reactor makes it possible to achieve the following:

1. to exchange plutonium in the two-component nuclear fuel cycle;
2. to increase the fraction of fissile isotopes to ~72% (~75% in case of refinement in BN-reactor MNUP fuel);
3. to compensate for the loss of its mass up to 220 kg/year when it is reused in VVER (up to 250 kg/year in MNUP fuel);
4. to expand the nuclear energy fuel base by 110 to 140 tons through involvement of plutonium from the SNF of RBMK UOX (uranium oxide fuel), unsuitable for the use in VVERs;
5. to reduce plutonium heat release and activity;
6. to passively burn up to 100 kg/year of Pu-241 – a source of Am (up to 115 kg/year in MNUP fuel).

Therefore, the development of a nuclear energy system from thermal and fast reactors with a joint fuel cycle will ensure the system’s long-term, reliable fuel supply. Moreover, this will allow the export capabilities of the State Atomic Energy Corporation ROSATOM to be expanded by providing the foreign customers with fuel supply services for thermal reactors (by improving the isotope vector of plutonium), balancing plutonium reserves in the global nuclear energy system, as well as at the back-end stage of the fuel cycle, by way of burning minor actinides.

The next function of sodium fast reactors is related to utilization of long-lived isotopes of minor actinides (Dekusar et al. 2019; Gulevich et al. 2020; Tuzov et al. 2023; Troyanov et al. 2025).

Neptunium is primarily generated in thermal reactors from uranium-235 (currently, the accumulation rate in the Russian nuclear power industry is ~200 kg/year); its generation in fast reactors (as a result of the decay of americium) is insignificant. Americium is generated in significantly larger quantities (~700 kg/year) (Dekusar et al. 2019), basically in the SNF of thermal reactors (as a result of the decay of the plutonium-241 isotope). It is also generated in fast reactors from the heavy isotopes of plutonium-241 and -242. From one BN-type reactor with homogeneous transmutation, depending on the fraction of americium involved, the curium production rate can reach 13 kg/year.

The reactor methods of MA utilization include homogeneous transmutation in the core fuel and heterogeneous transmutation in specialized assemblies of the fast neutron reactor’s radial blanket.

The efficiency of homogeneous transmutation of neptunium and americium depends on the fraction of MA in the fuel and on the isotopic composition of plutonium in recycling. The maximum fraction of homogeneously involved MA is determined by the heat release of the fresh and spent fuel assemblies, respectively. The maximum fraction of MA in the fuel in case of multiple recycling of americium and neptunium separately is 2.2% and 2.8%, respectively. Moreover, the transmutation rates will be equal to ~50 and 100 kg/year. The disadvantages of the homogeneous method are its low transmutation efficiency, which is ~20–30%, deterioration of the isotopic composition of plutonium in recycling (increase in the fraction of Pu-238 and Pu-242), and high rates of curium accumulation (in americium transmutation).

Heterogeneous transmutation of americium by way of its dilution with inert material makes it possible not only to achieve a burning rate of ~ 80 kg/year, but also a significantly higher efficiency of ~ 90%. Another advantage of this method is the feasibility of burning curium, a product of americium transmutation that is required for the complete closure of the minor actinides fuel cycle.

After irradiation, the specialized assemblies are suggested to be held in the boron shielding of the reactor and in the external storage facility for ~60 years in order to eliminate the need for handling high-level “hot” curium-244. Within this time, the amount of curium will decrease 3 times (only low-level long-lived isotopes -245 and -246 will remain), and the activity and heat release of the irradiation devices (ID) will decrease to almost the level of “fresh” IDs containing americium, suitable for reprocessing by means of the currently available technologies.

The study has demonstrated that the least expensive option in terms of the volumes of Am and Cm circulation and handling consists in the option when “clean” fuel is used in the core, and the “native” americium produced in the core is irradiated together with the “external” americium in the IDs with the radial blanket moderator of the BN-reactor for 8 years. In this case, the efficiency of Am transmutation is ~90%, and handling of americium and curium after long-term storage of IDs for 60 years is minimal (Fig. 4). After their transmutation and storage, a small amount of remaining low-level Am and Cm isotopes can be re-directed without their separation for “afterburning” in the radial blanket.

Figure 4. 

Diagram of the fuel cycle to burn Am in the BN-reactor radial blanket.

The economic efficiency growth in this option is associated with a decrease in the radiation load on the entire fuel cycle of the reactor, as well as with almost complete elimination of multiple Am recycling and “hot” Cm handling, afterburning of the “unseparated” mixture of americium and curium in the IDs of the radial blanket, and elimination of a curium storage facility.

The final function of the BN-type reactors is related to the production of industrial and energy-generating isotopes. The most well-known of them, such as cobalt-60 (used for sterilizing products and for medical purposes) and plutonium-238 (used in self-contained energy sources) can be efficiently produced in the radial blanket of fast reactors.

The research has been conducted where the production of cobalt is considered in the second row of the BN-800 radial blanket (the first row is occupied by steel shield assemblies to protect the core from thermal neutrons). The required spectral parameters in irradiation devices of a fast reactor are primarily achieved through the use of metal hydrides. In Russia, there is a positive experience in the use of zirconium hydride for Co-60 production in a BN-type reactor. Within one year of irradiation, the BN-800 power reactor allows cobalt-60 to be produced up to 8 MCi with a specific activity of ~100 Ci/g (Varivtsev et al. 2014; Risovanyi 2020; Dzugkoeva et al. 2021; Shaginyan and Korobeinikova 2023), which, in turn, is at least 5 times faster than in RBMK reactors (Borshchev et al. 2003; Bazulin et al. 2025). This isotope is not produced in VVER reactors.

A specific feature of the target isotope Pu-238 is the presence of only α-radiation with thermal power of 570 W/kg, therefore it can be used as a heat source near electronic devices and allows for manual operations. Until now, this isotope has been produced in a commercial graphite reactor. The production of Pu-238 in the radial blanket of a fast reactor without the use of a dilution technology does not allow for the production of Pu-238 of the required quality. However, the studies have demonstrated the feasibility to produce Pu-238 of the required quality from Np-237 in special irradiation devices in the radial blanket of a fast reactor. With application of this method in BN-type reactors, the annual production of Pu-238 is over 10 kg per year. The production rate of plutonium-238 in the existing specialized reactors, as well as in the research in thermal power reactors, does not exceed 3 kg/year (PNNL 1999; Daily and McDuffee 2020; Shmelev et al. 2020; Shmelev et al. 2022).

A stepwise transition of Russian nuclear power industry to a two-component nuclear energy system with thermal and fast reactors in a single closed nuclear fuel cycle makes it possible to solve a number of well-known deferred problems of the current power industry. Besides, fast reactors in the two-component NES can undertake a number of functions, such as time-unlimited fuel supply to the system, and its cleaning from long-lived and hazardous waste, and production of important radioisotopes.

The preliminary estimate of these additional, in relation to electricity generation, capabilities demonstrates the following:

1. Sodium fast reactors have the required neutronic potential to triple the installed nuclear capacities by the end of the century and to develop and construct the facilities for reprocessing their spent nuclear fuel.
2. The breeding properties of a BN-type reactor with MOX (and, in the future, MNUP) fuel ensure the management of plutonium production from its “burning” at a rate of ~40 kg/year to an excess breeding of ~250 kg/year, with its importance being increased. In addition, with a high breeding ratio in “clean” fuel, the production rates of the native americium and curium are minimal: ~ 23 and 0.6 kg/year, respectively.
3. The function of “refining” the plutonium from MOX SNF of thermal reactors allows for arrangement of a joint fuel cycle, where one sodium fast reactor will provide fuel for 1 ÷ 2 VVER-S-600 or VVER-SKD reactors.
4. The least expensive NFC option in terms of the volume of Am and Cm recycling and handling, is the option, where “clean” fuel is used in the core, and the native americium produced in the core is irradiated together with the “external” americium in the IDs with a moderator in the radial blanket of a BN-type reactor. In this case, the efficiency of Am transmutation is ~90%, and americium and curium handling after 60 years of IDs’ long-term storage is minimal, and the storage facility for curium can be eliminated.
5. In the radial blanket of BN-type fast reactors, it is possible to produce the Co-60 and Pu-238 isotopes, which are highly-demanded in the industry and medicine, in the amount of up to 8 MCi/year and over 10 kg/year, respectively.

All this gives grounds for the creation of a multi-functional fuel component for the future two-component NES based on sodium fast reactors, which could significantly improve the economy of Russia’s nuclear energy system.