Analytical studies on the potential of 238Pu production in the large-size fast neutron power reactor

Analytical studies on the potential of 238Pu production in the large-size fast neutron power reactor Ruslan A. Vnukov1, Elena A. Rodina1, Yurii S. Khomyakov1 1 JSC Proryv, Moscow, Russia Corresponding author: Ruslan A. Vnukov ([email protected]) Academic editor: Osama Ashraf ♦ Received 1 September 2025 ♦ Accepted 9 December 2025 ♦ Published 16 December 2025 Citation: Vnukov RA, Rodina EA, Khomyakov YS (2025) Analytical studies on the potential of 238Pu production in the large-size fast neutron power reactor. Nuclear Energy and Technology 11(4): 293–299. https://doi.org/10.3897/nucet.11.170677 Abstract The possibility of obtaining 238Pu in the BN-1200M (about 1,200 MWe) fast neutron power reactor is considered. To obtain products with a purity of 80–85% in terms of 238Pu, options were proposed with installation of irradiation assemblies (IA) with breeding and moderating elements in the radial blanket. Neptunium and americium oxides and their mixtures were considered as target materials of the breeding elements for producing target nuclide. Zirconium hydride was considered as a moderator. The effective density of targets, IA arrangement, target/moderator materials content ratio, and irradiation time were varied. Calculations were made to determine mass and composition of produced plutonium and IA decay heat, as well as IA effect on the global and local pin-to-pin power profile in the standard fuel assemblies of the reactor core. When using targets made of neptunium oxide, options were effective only at an effective density of approximately 1.1 g/cm³ were only effective because of self-shielding effect. Irradiation during 4–5 years ensures required 80% purity of 238Pu with target purity in terms of 236Pu within 2 ppm. When using Am, 236Pu purity is also ensured, but 238Pu fraction is somewhat lower than 80% due to the higher fraction of 242Pu. The use of Am and Np mixture makes it possible to reduce the fraction of 236Pu and prevent high accumulation of 242Pu, but in this case 242Cm accumulation should be taken into account, and appropriate time period for its decay heat reduction should be chosen. Keywords Radioisotope products, BN-1200 fast reactor, irradiation device, moderator element, neptunium, americium, zirconium hydride Introduction The demand for radioisotopes is constantly growing, this leading to a need for higher loading of existing research reactors (Gadzhiev et al. 2001; Kotchnov 2011). This is not always compatible with the main functions of such facilities (Klinov et al. 1989). If so, the cost of nuclear fuel and reactor operation should also be included in the cost of isotopes, which can significantly increase their cost. The creation of specialized isotope production reactors is possible and is being considered, however, it is associated with rather high costs of the construction and operation of nuclear installations, which are difficult to recover by the radioisotope products. In this regard, it is advisable to consider the possibility of producing radioisotope products in power reactors along with the main process, i.e., nuclear energy production and/or SNF reprocessing in a closed nuclear fuel cycle. Copyright Vnukov RA et al. This is an open access article distributed under the terms of the Creative Commons Attribution License (CC-BY 4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original author and source are credited. Nuclear Energy and Technology 11(4): 293–299 DOI 10.3897/nucet.11.170677 Research Article Vnukov RA et al.: Studies on the potential of 238Pu production in the fast neutron reactor294 Then it can be assumed that the cost of isotopes will be relatively small and related only to the adaptation of reactor technology elements for this task. The first attempts to organize the production of isotopes in NPP power reactors were made in the 1980–90ies in relation to RBMK reactors and BN-350 and BN-600 fast reactors (Zvonarev et al. 1994; Evdokimov et al. 1997; Mal’tsev et al. 1999), primarily in relation to the isotope 60Co. Currently, industrial production of 60Co is organized in RBMK reactors (Borshchev et al. 2003; Pereguda et al. 2013). However, no further construction of such reactors is envisaged, and the shutdown of all existing RBMK reactors is planned in the next 10 years. Work has been carried out on fast reactors to develop technologies for the production of 60Co and a number of other isotopes, such as 89Sr for medical purposes (Gadzhiev et al. 1997; Zvonarev et al. 1997) and 51Cr and 37Ar for scientific purposes (Zvonarev et al. 1996). Recently, the topic of producing radionuclides for energy purposes (nuclear batteries) has become relevant again (Shaginyan et al. 2022). The choice of technology depends on the level of required activity, service life and pricing policy. Isotopes 90Sr and 238Pu have turned the most popular. Plutonium isotope 238Pu is mainly used in radioisotope electric generators (RITEG) designed for power supply of spacecrafts and remote autonomous installations, as well as in significantly smaller numbers in cardiostimulators. 238Pu has a long half-life (87.7 years) and high specific energy release (570 W/kg) in the absence of hard gamma radiation, which makes this isotope suitable for the longterm steady state power supply. There are some limitations related to the composition of produced plutonium. Regarding the process of producing 238Pu in nuclear reactors, there are two important requirements for its composition: 238Pu fraction is at least 80% (Kruglov 1985) (in some publications - at least 85% (Kulikov et al. 2023)), and 236Pu fraction is within 2 ppm. For medical needs, 236Pu fraction should not exceed 3 × 10-7 (Gromov 1983). The undesirability of 236Pu is caused by the formation of daughter decay products (232U) and a source of hard g radiation (208Tl). Production of 238Pu in Russia is currently carried out mainly at Mayak Industrial Association, which is the largest Russian producer of isotopes (Ionizing Radiation Sources and Compounds 2020). The bulk of radioisotope products of Mayak are exported, accounting for up to 20% of the global isotope production. To obtain 238Pu with the target parameters, a technology was developed for extracting neptunium from spent nuclear fuel and irradiating 237Np in thermal neutron reactors. The potential increase in demand for 238Pu, including that due to the growing number of space programs in various countries (U.S. Department of Energy 2010), has now led to the search for alternative ways of its production. The goal of the work is to develop and calculate the feasibility of 238Pu production in a fast neutron power reactor (FR) and adapt its core for this task (BN-1200M (Vasiliev et al. 2021) was chosen as the base for this work). To achieve this goal, an analysis of potential methods of this isotope production, the configuration and arrangement of specialized irradiation devices was carried out, as well as computational analysis of the irradiation parameters and characteristics of the isotope being produced, and an assessment of the effect on the core neutronics in order to maintain them within acceptable limits. Description of calculation models A preliminary analysis showed that 237Np irradiation in the traditional fast neutron spectrum of FR leads to the formation of 236Pu as a result of the threshold reaction (n,2n) in an unacceptable amount. Therefore, in order to produce 238Pu, an approach was adopted that was previously used for 60Co production, i.e., creation of special irradiation devices (ID), with a moderator based on zirconium hydride, placed in the radial blanket of the BN-1200M. For computational analysis, several ID designs were developed based on a standard BN-1200M radial blanket fuel assembly containing 127 elements with 14 mm diameter, 1 mm thick cladding. Similar approaches are used, for example, in (Yusupova et al. 2022; Kashirina et al. 2025). Some of the elements, which are hereinafter referred to as moderator elements (MEL), were filled with zirconium hydride, ZrH1.85, and the remaining elements (hereinafter referred to as breeding elements - BEL) were filled with target material, which was considered to be neptunium or americium oxides, or their mixtures of various densities. The considered designs of irradiation assemblies containing different numbers of MEL and BEL are shown in Fig. 1, and ID parameters are presented in Table 1. To reduce the impact of MEL on the fuel elements of the peripheral fuel assemblies, BEL must be placed on the periphery of the ID. Quite a number of IDs (84) were placed in the first row of the BN-1200M radial blanket, and the cross-section of the BN-1200M reactor model is shown in Fig. 2. To assess energy release nonuniformity, peripheral assemblies of the core are assumed to be heterogeneous. Analytical studies were carried out to assess the possibility of meeting the requirements for commercial purity of 238Pu and preventing maximum values of fuel Table 1. ID parameters Parameters Values BEL material NpO2, AmO2, NpO2 - AmO2 mixture (0.1:0.9 to 0.9:0.1 ratios) Target effective density, g/cm30.3–11.1 for NpO2 0.3–10 for AmO2 0.2–10 for NpO2 - AmO2 mixture BEL and MEL diameter, mm 14 × 1 BEL target height, cm 100 Irradiation time, eff. days Up to 3,600 (up to 11 reactor runs, 315 eff. days each) Nuclear Energy and Technology 11(4): 293–299 295 element energy release from exceeding values for a standard core. Two types of disturbances of the BN-1200M neutron field caused by ID should be noted, namely: global disturbance throughout the core (a change in the distribution of the neutron flux) and local disturbance in the area of peripheral core fuel assemblies bordering ID (disturbances caused by a change in the neutron spectrum). Significant unevenness of the power profile in the fuel elements of peripheral fuel assemblies, as shown in Fig. 3, should be noted. Max power value (~15 kW) is reached in the fuel elements located closer to the core center (farther from the disturbance source, i.e., ID with ZrHx). This value was taken as a reference allowing exclusion of the necessity of changing core and peripheral fuel assemblies design. Results and discussions From the results of studies on various options of changes in the effective density, number of rows and arrangement of BELs, and irradiation time, only some of them are selected and discussed below to reflect the general specifics. Using neptunium target Produced plutonium mass as a general function of NpO2 effective density is shown in Fig. 4 for ID design containing 1 row of BELs/91 MELs (irradiation time 1,440 eff. days, post-irradiation storage time 3 years). An increase in the fuel material density leads to an increase in 236Pu content. This is due to the spatial blocking effects leading to a decrease in the medium-volume density of thermal and near-thermal neutrons, and as a result, an increase in the relative fraction of fast neutrons. Based on this, a conclusion can be made that the use of effective Figure 1. Cross-sections of ID models. Figure 2. Cross-section of BN-1200M reactor model with 84 IDs. Figure 3. Pin-to-pin distribution of power (kW) in peripheral fuel assemblies of the initial BN-1200M model (without ID). Vnukov RA et al.: Studies on the potential of 238Pu production in the fast neutron reactor296 densities higher than 2 g/cm3 is impractical from the point of view of material purity. However, it should be noted that density decrease leads to a decrease in 238Pu fraction with an increase in 239Pu fraction. Therefore, decrease of the effective density below 1 g/cm3 for the irradiation time range of 1,440–2,880 eff. days is also impractical due to the high content of heavy isotopes of plutonium. An increase of BELs number with a corresponding decrease of MELs number generally leads to a deterioration of 238Pu quality in terms of 236Pu isotope with an increase in the total mass of produced plutonium and 238Pu fraction in it, i.e., it has the same trends as in case of density increase. Increasing BEL rows number from 1 to 2 or 3 rows leads to an increase in 236Pu fraction from 3.7 ppm to 4.1 ppm and 10 ppm, and 238Pu fraction - from 85% to 88% and 95%, respectively. Mass of accumulated Pu increases by 54% and 80% with an increase in the total loading of Np oxide by 83% and 233%, respectively. In order to reduce the effects of blocking, alternative options with uniform arrangement of BELs and MELs (Alter-1 and Alter-2, respectively) were considered. In terms of BELs number, the configuration is similar to the option with two BEL rows around MEL in ID. The results of plutonium production for Alter-2 are presented in Table 2. Alter-1 option values are quite similar, however, the use of Alter-2 option leads to lower power of fuel pins in peripheral fuel assemblies and decay heat rate of ID (Table 3). Out of the large number of options considered, four most suitable options have been identified, in which 238Pu isotopic composition quality parameters can be achieved with 1.1 g/cm3 accepted density of the target material with NpO2. Generalized estimates of production parameters to ensure 238Pu fraction equal to at least 85% and 80% are given in Tables 3, 4, respectively. Based on the integral assessment of accumulation in the ID, extrapolations were made to determine the annual reactor capacity when the 1st row of the radial blanket is fully filled with 84 IDs. An estimate has also been made of the required average annual number of simultaneously irradiated IDs to achieve a traditional capacity of about 3 kg/year. The results show the potential of producing 238Pu in a 1,200 MWe reactor in large amount exceeding the existing estimates of its world current consumption. Despite the local power peak, the fuel element power in the peripheral fuel assemblies of the proposed configurations does not exceed the maximum values for the initial core design with the standard radial blanket. The linear power on the fuel element in the core as a whole does not exceed 500 W/cm in all selected options. ID installation leads to a decrease in the reactivity by -0.6% Δk/k, and as the assemblies are irradiated, the effect decreases to -0.4%. The loss of reactivity can be eliminated by changing the fuel plutonium enrichment in the standard fuel assemblies. Using americium target Americium, as well as neptunium, is a minor actinide, the incineration of which. On the one hand, it is a vital task of fast reactors. On the other hand, it is a potential source of 238Pu, which is the main nuclide related to the transmutation Figure 4. Mass of produced Pu vs NpO2 effective density: (a) – 238Pu, (b) – 236Pu. Nuclear Energy and Technology 11(4): 293–299 297 of 241Am. The key features of producing 238Pu from 241Am are the absence of a direct channel for 236Pu formation, this, of course, being an obvious advantage, and a more significant accumulation of 242Pu (at least 17%) to dilute the target isotope. Calculations show that maximum 238Pu fraction is about 76%, although 236Pu fraction does not exceed 1 ppm. This is due to decay pattern of 242Am, which is a daughter product of the radiation capture reaction by 241Am. According to it, 242Cm is produced with a probability of 82.7%, decaying with a half-life of 162.8 days to 238Pu and with a probability of 17.3% - to 242Pu. In this case, 238Pu can also be produced in a fast neutron spectrum, but due to the relatively low neutron flux in the radial blanket area the moderator increases the effectiveness of the process, i.e., the specific production rate per target mass unit as shown in Table 5. However, the moderator decreases 238Pu fraction to 64–65% by accelerating its transmutation to 239Pu. Another feature of this 238Pu production method is the increased ID decay heat caused by 242Cm. It can reach 3.5–20 kW after 2 years of storage, depending on the mass of the loaded americium. This may require longer ID storage time after irradiation. Using target made of NpO2 and AmO2 mixture The combined use of americium and neptunium oxides was studied to obtain a synergistic effect from the advantages of individual components, namely: required high 238Pu content, typical for Np, and low 236Pu content inherent in Am target. In fast neutron spectrum, in the absence of a moderator, it is possible to obtain 238Pu purity up to 79% with a 1.9 ppm 236Pu fraction and 18% 242Pu fraction with a ratio of Np/Am = 1/9 and irradiation time of 2,880 eff. days. As regards two-band ID model with BELs located on the periphery and MELs located in the center, it is possible to ensure the required purity in terms of 236Pu and about 80% 238Pu fraction with low densities (0.2 g/cm3). In this case, it is necessary to increase neptunium fraction in order to maintain 238Pu fraction equal to 80%. It is recommended to adopt Np/Am = 1.0/1.1 ratio typical for VVER SNF after its limited storage during of 7–8 years. The results of calculations of various ID options with ZrHx are given in Table 6. As regards the unlocked ID options (Alter-1 and Alter-2), the required 238Pu quality can be reached with a fuel material density of 1 g/cm3. The two options are similar in terms of parameters and, in terms of produced plutonium mass, are ahead of the symmetrical ID options discussed above. The dynamics of changes in the operating time of 238Pu is shown in Table 7. For given power density of the fuel element, power values of the core fuel assemblies are close to those for ID option with Np density of 1.11 g/cm3 and do not exceed 15 kW. Table 2. Results of plutonium production for Alter-2 ID with NpO2 target with 1.1 g/cm3 eff. density Irradiation time, eff. days Fraction, ppm Plutonium isotope composition, % Pu mass, kg/ year 236Pu 238Pu 239Pu 240Pu 241Pu 242Pu 360 4.3 / 3.9 94.8 4.7 0.3 0.1 0.0 66 720 3.6 / 3.4 91.2 7.5 0.9 0.5 0.0 62 1,080 3.0 / 3.1 88.4 9.1 1.4 0.9 0.1 58 1,440 2.7 / 2.8 86.1 10.3 2.0 1.4 0.2 53 1,800 2.7 / 2.6 84.2 11.2 2.5 1.8 0.4 49 2,160 2.3 / 2.4 82.4 11.8 3.0 2.2 0.6 45 2,520 2.0 / 2.0 80.8 12.3 3.6 2.6 0.8 41 2,880 1.9 / 2.0 79.2 12.7 4.1 3.0 1.0 37 Table 3. Parameters of plutonium production with limited 238Pu fraction (85%) Characteristics \ ID options 1 BEL row 2 BEL rows Alter-1 Alter-2 Irradiation time, eff. days. 1,440 1,800 1,800 1,440 Storage time, years 3 3 2 2 236Pu fraction, ppm 1.8 1.8 1.6 1.7 238Pu fraction, % 85 86 85 86 Production rate in the reactor, kg/year 27 39 52 53 Average annual ID number for production rate about 3 kg/year 9.3 6.7 4.9 4.7 Max power of fuel element in the peripheral fuel assembly, kW 15.7 11.4 14.3 11.8 Max linear power of fuel element, W/cm 484 488 480 490 ID decay heat, kW 1.42 2.15 3.12 2.98 Reactivity loss at the beginning of the core lifetime, Δk/k % -0.5% -0.5% -0.6% -0.6% Table 4. Parameters of plutonium production with limited 238Pu fraction (80%) Characteristics \ ID options 1 BEL row 2 BEL rows Alter-1 Alter-2 Irradiation time, eff. days. 2,160 2,880 2,880 2,520 Storage time, years 3 3 0 0 236Pu fraction, ppm 1.3 1.7 1.9 2.0 238Pu fraction, % 81 82 81 81 Production rate in the reactor, kg/year 23 32 39 41 Average annual ID number for production rate about 3 kg/year 11.2 7.9 6.4 6.2 Vnukov RA et al.: Studies on the potential of 238Pu production in the fast neutron reactor298 After irradiation for 1,440–1,485 effective days and post-irradiation storage for 30 days ID decay heat ranges from 50.9 to 52.1 kW. Thus, when using Am in IDs, it is necessary to store them, as the standard core fuel assemblies, in the in-vessel storage or boron shielding rod cells to reduce ID decay heat level. Conclusion Analysis of the possibility of combining the solution of two problems, namely: transmutation of long-lived MA (Np, Am) and production of heat-generating radioisotope, 238Pu, has shown that, in accordance with the requirements for its purity, this is potentially possible with a heterogeneous transmutation pattern using moderator in the radial blanket. Production parameters and isotopic composition of produced plutonium depend significantly on the design of irradiation device, i.e., moderator/target material volume ratio (MEL and BEL numbers), density of the target material, patterns of BEL and MEL arrangement in the ID and irradiation time. After analyzing and optimizing a large number of options, four ID designs with NpO2 are proposed ensuring 238Pu production time in accordance with the requirements for 238Pu content (at least 85%) and 236Pu content (within 2 ppm) with a total production rate of ~25 to ~50 kg/year for full loading of the first row of the BN-1200M radial blanket. Np density in the target should not exceed 1.1 g/ cm3. To decrease 236Pu fraction to the value not exceeding 2 ppm, it is necessary to store ID with 85% purity in terms of 238Pu for 2–3 years before reprocessing, and there is no need for storing in case of 80% purity. These potential production volumes may and most likely actually exceed the world demand for this isotope; to achieve realistic requirements of 3 kg/year, 238Pu will require installation of 5 to 10 ID in the radial blanket of the BN-1200M.ID presence may cause an increase in power of the core peripheral fuel assemblies adjacent to ID. However, in spite of the disturbance of the power profile at the core boundary in the proposed options, the fuel element power does not exceed the maximum values of 15 kW for standard fuel elements, and the linear power of the hottest fuel pins of the core is within 500 W/cm. Calculations of decay heat values for proposed ID options after storage during 30 days gave a value of about 3 kW, this making possible their unloading taking into Table 5. Parameters of production for max 238Pu fraction with AmO2 target Characteristics \ ID options 2 BEL rows 3 BEL rows 5 BEL rows No MEL Effective density, g/cm30.7 0.7/7 0.7/7 7 Irradiation time, eff. days 990/2,880 990/1,440 1,080/1,800 1,800/2,880 236Pu fraction, ppm 0.0/0.0 0.0/0.3 0.1/0.5 0.8/1.1 238Pu fraction, % 73/64 74/77 75/77 77/76 Pu fraction, % 22/47 18/5.4 11/6.0 6.0/9.9 Production rate in reactor, kg/year 31/26 34/71 26/83 85/102 Average annual ID number for production of about 3 kg/year 8.1/9.7 7.4/3.5 9.7/3.0 3.0/2.5 *storage time is not indicated, since none of the options meets purity requirements in terms of 238Pu. Table 6. Parameters of 238Pu production in case of combined use of Am and Np Characteristics \ ID options 1 BEL row 2 BEL rows Alter-1 Alter-2 Irradiation time, eff. days 2,160 2,880 2,880 2,520 Storage time, years 3 3 0 0 236Pu fraction, ppm 1.3 1.7 1.9 2.0 238Pu fraction, % 81 82 81 81 Production rate in the reactor, kg/year 23 32 39 41 Average annual ID number for production of about 3 kg/year 11.2 7.9 6.4 6.2 Characteristics \ ID options 1 BEL row 2 BEL rows 4 BEL rows Irradiation time, eff. days 360 360 1,080 236Pu fraction, ppm 1 2 1.7 238Pu fraction, % 78 81 80 Pu fraction, % 18 17 32 Production rate in the reactor, kg/year 19 37 38 Average annual ID number for production of about 3 kg/year 13.3 6.8 6.6 *ID storage time for presented options is 0 years Table 7. Results of plutonium production for Alter-1 and Alter-2 ID Irradiation time, eff. days Fraction, ppm Plutonium isotope composition, % Pu mass, kg/year 236Pu 238Pu 239Pu 240Pu 241Pu 242Pu 360 2.8 84.2 2.7 0.3 0.1 12.8 51.5 720 2.5 83.7 5.0 0.6 0.3 10.5 55.5 1,080 1.9 82.1 7.1 0.9 0.6 9.3 55.2 1,440 1.8 80.4 8.7 1.4 1.1 8.5 52.2 1,800 1.6 78.7 10.0 1.9 1.6 7.9 48.4 2,520 1.1 75.2 11.7 3.4 2.6 7.0 45.6 2,880 1.0 73.5 12.2 4.3 3.2 6.8 39.0 Nuclear Energy and Technology 11(4): 293–299 299 account similar values for standard fuel assemblies or those with homogeneous use of MA. It has been demonstrated that in case of using AmO2 as a target material, the resulting 238Pu contains no more than 1 ppm of 236Pu. However, 238Pu isotope fraction does not reach 80%. When irradiating NpO2 - AmO2 mixture, it is possible to obtain 238Pu fraction above 80% and low content of 236Pu (about 1–2 ppm), but it requires ID storage after irradiation, as in the case of standard fuel assemblies. 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