The VVER-1200/AES-2006 is recognized as a leading Gen III+ nuclear reactor design, meeting stringent international safety standards. This study evaluates the use of a novel PuO₂-ThO₂ duplex fuel, incorporating weapon-grade plutonium (WgPu) and thorium, for a hypothetical VVER-1200 assembly. The research also explores incorporating minor actinides (MAs) for transmutation, comparing two methods: MAs coated on WgPuO₂ and MAs mixed with WgPuO₂ as integral fuel burnable absorber rods. Neutronic properties of these fuels are compared to those of LEU-fueled assemblies. The results show a 135% higher burnup for the duplex fuel compared to LEU, with extended criticality, reduced reactivity swings, and lower Pu-239 concentrations upon discharge. While Np-237 and Am-241 concentrations decrease, Am-243, Cm-244, and Cm-245 increase, but overall radiotoxic waste is reduced. Enhanced safety coefficients are also observed, within acceptable LWR ranges.

VVER-1200 assembly, Duplex Fuel, Plutonium, Thorium, Minor Actinides

Recently, there has been an increase in global demand for energy to achieve social and environmental justice for the world’s growing and developing inhabitants. Nuclear energy appears to be an appealing option due to its low carbon footprint; however, it has faced many challenges since its inception, manifested in a lack of trust from the public, anti-nuclear, and political classes for a variety of reasons, most notably the operational safety and nuclear wastes in spent fuel. To reestablish this lost trust, nuclear reactors planned for construction or nuclear fuels proposed for use must overcome the aforementioned obstacles.

The use of conventional nuclear reactors, for example, generates large amounts of irradiated fuel (i.e. spent fuel). A conventional light water reactor (LWR) with a capacity of 1 GWe that uses low-enriched uranium oxide (LEU UO₂) will typically discharge 20 to 25 tonnes of irradiated fuel per year of operation. Irradiated UO₂ fuel is composed mainly of uranium, predominantly U-238, with a small fraction of unburned U-235, which typically accounts for about 4 to 5 wt.% of the spent fuel. Additionally, it contains approximately 1 wt. % of plutonium and around 0.1 to 0.2 wt.% of MAs, along with about 4 to 5 wt.% of other fission products (OECD/NEA 2013). Given the above, large amounts of plutonium have accumulated in the world in the form of burned fuel. According to IAEA reports, the ongoing global spread of discharged plutonium poses a significant proliferation risk (IAEA 1998; Kotlyar et al. 2017). Furthermore, fuel with a low burnup, typically around 1 GWd/ton, has been shown to yield weapons-grade plutonium (World Nuclear Association 2017; Abdelghafar Galahom 2020). Because of growing public and political concern about the potential misuse of this plutonium, the need to address this surplus has gained prominence. Therefore, an appealing and viable approach for reducing plutonium inventories is the reutilization of this material as reactor fuel. This strategy not only helps mitigate the proliferation risk but also offers a practical solution for managing surplus plutonium.

Long-lived byproducts, commonly referred to as MAs, are produced through activation and fission during reactor operation and tend to accumulate in spent fuel. The most commonly abundant MAs in this context typically consist of a triad of elements, namely neptunium (Np), americium (Am), and curium (Cm). As the used fuel is subjected to higher burnup levels, additional minor actinides such as berkelium (Bk) and californium (Cf) also become significant contributors to the mix. MAs pose a problem to nuclear environmental safety because they are responsible for a significant contribution to spent nuclear fuel radiotoxicity (OECD/NEA 2013) having major consequences in radiation protection issues of waste treatment and disposal. To reduce the amount of actinides in spent fuel, it is best to recycle them alongside the plutonium in thermal and fast reactors. Np-237, Am-241, and Am-243 have very large thermal neutron capture cross-sections (Fig. 1), therefore, these MAs can capture thermal neutrons and transform into Pu-239, Am-242, Am-244, Cm-243, or Cm-245, which have thermal fission cross-sections between large and moderate, and thus may serve as alluring fissile materials (Liu et al. 2014; Galahom and Sharaf 2021). Furthermore, since MAs have very large thermal neutron capture cross-sections, they can act as burnable absorbers (BAs) when introduced into reactors to reduce the initial reactivity of fresh fuel (Benrhnia et al. 2022a). Consequently, the use of these MAs may be beneficial in reducing initial reactivity due to the presence of highly fissile Pu-239. The epithermal cross-sections of these MAs contain resonance regions, which are beneficial in the LWRs spectrum to increase the negative temperature coefficients, particularly the moderator temperature coefficient, because in the case of Pu-based fuels, as the temperature of the moderator increases, the Maxwellian region of the neutron spectrum shifts slightly towards higher neutron energies, potentially benefiting the fission resonance of Pu-239 and Pu-241 (Richards and Serfontein 2014; Kabach et al. 2021; Zou et al. 2021). What’s more, one of the distinct benefits of recycling MAs and using them as a BAs option is that it may be a way to reduce the concentration of boric acid in the moderator. Therefore, all of these characteristics can significantly reduce the possibility of a criticality accident in a fuel cycle (Liu et al. 2014).

Figure 1. 

Capture cross-sections of MAs (Zou et al. 2021) (a); Th-232 and U-238 (b); and Pu-238, Pu-240, and Pu-242 (c).

According to the available literature, many studies have been conducted to investigate the recycling of plutonium and MAs separately or together in thermal or fast reactors. However, it appears that the vast majority of studies conducted by researchers in LWRs are focused on the recycling of plutonium in uranium-based mixed-oxides (Emmett and Ans 2000; NEA 2002; Gomin et al. 2006) or thorium-based fuels (Galahom 2018; Galahom 2020). The majority of studies on MAs suggest recycling in LEU-fueled reactors (Taiwo et al. 2006; OECD/NEA 2013; Liu et al. 2014), and no research study has focused on using both plutonium and MAs in thorium-based fuel. Burning plutonium and MAs in LEU-fueled LWRs is ineffective because the fertile U-238 is the main source of the residual high quantities of plutonium and other long-lived radioactive actinides. On the other hand, replacing U-238 with Th-232 will result in not only more new fissile material (U-233) but also lower waste inventories. U-233, like Pu-239 and Pu-241, is a fissile isotope with a long half-life that can be produced in reactors via single-neutron capture. It has a fast critical mass that is nearly identical to Pu-239, but the spontaneous fission rate is much lower, reducing the problem of a spontaneous fission neutron prematurely initiating the chain reaction to negligible levels (Kang and von Hippel 2001). Another advantage is that, unlike U-233, the plutonium produced from standard UO₂ fuels is a wasting asset due to the fissile Pu-241’s short half-life, which decays to non-fissile Am-241 (Galahom 2018).

Another significant advantage of incorporating Th-232 is its notable sensitivity to resonance cross-sections concerning temperature under self-shielded conditions, known as the thorium Doppler effect. Although the effective resonance integral for a given temperature is greater for U-238 when compared to Th-232, the change in this integral per degree is more pronounced for Th-232. Consequently, reactors utilizing Th-232 will exhibit more substantial negative feedback on neutron multiplication with increasing fuel temperature (Doppler coefficient) than those using U-238 (Belle and Berman 1984; Kabach 2021; Kabach et al. 2021a, 2021b). This characteristic results in an enhanced negative Doppler coefficient. Consequently, it is proposed that WgPu and MAs be recycled in thorium-based reactors to breed additional nuclear fuel and provide a solution for nuclear waste management.

There are three main approaches to utilizing thorium in LWRs: homogeneous, macro-heterogeneous, and micro-heterogeneous. The homogeneous method, which mixes thorium with fissile material (e.g., plutonium in this case), is the simplest. The resulting fuel is called mixed oxide fuel (Abdelghafar Galahom and Ibrahim 2022; Abdelghafar Galahom et al. 2024). However, this method increases fuel costs over the cycle. It also requires the fabrication of mixed oxide fuel, and the fuel exhibits suboptimal neutronic performance due to the higher fissile loading needed, compared to LEU, to maintain the standard cycle length in PWRs (Li et al. 2021; Benrhnia et al. 2022).

If thorium is separated from the fissile fuel, i.e., fissile fuel in the seed region of a fuel core or assembly, and thorium in the blanket region, the approach is known as macro-heterogeneous (i.e. seed-blanket) (Bromley 2016; El Banni et al. 2024). The concept of a macro-heterogeneous scheme can effectively use neutrons for breeding; however, this concept causes power imbalance and increases the temperature of the seed region at the beginning of cycle (BOC) (Li et al. 2021). The micro-heterogeneous approach has three main design types. The first method involves alternately stacking fissile and thorium pellets within a fuel rod in the axial direction. The second is a fuel assembly with fissile and thorium fuel rods arranged in a checkerboard pattern. The final type, known as duplex pellets, contains thorium and fissile zones radially with different volume ratios. According to relevant studies, the burnup of the MIT-proposed duplex fuel design can achieve a longer discharge life with a single batch for once-through fuel management (Joo et al. 2004; Rabir et al. 2020; El Kheiri et al. 2023a, 2023b).

The current work was undertaken with the primary objective of assessing the effectiveness of WgPuO₂-ThO₂ duplex fuel with MAs in an assembly-level analysis. The first objective of this research is to compare the fuel cycle performance of WgPuO₂-ThO₂ duplex fuel in the hexagonal assembly of the VVER-1200 reactor to that of a conventional LEU assembly. The next important objective of this research is to investigate the effects of adding MAs as the form of integral fuel burnable absorbers (IFBAs) located symmetrically within the fuel assembly on the WgPuO₂-ThO₂ duplex case, to investigate their effect on the fuel cycle, neutronic performance, and nuclear fuel depletion behavior. Because IFBA may contain neutron-absorbing materials that can be homogeneously mixed with fuel rod components or coated with the fuel rod, both options were explored in this research. Furthermore, the characteristics of the VVER-1200 assembly have been investigated using various types of loading designs and different loading thicknesses with their equivalent concentrations.

In the current analysis, calculations were conducted with the absence of soluble boron in the moderator (0 ppm), and all control rods were fully withdrawn. The simulations were carried out at the temperatures outlined in Table 3, maintaining a constant power density of 40 kW/kg for all cases. It’s noteworthy that in these neutronic calculations, the outer surfaces of the assemblies were treated as reflective, indicating the absence of neutron leakage and the eigenvalues will be considered infinite multiplication factor (k_inf). However, in a typical large pressurized water reactor, such as a VVER, there’s usually about a 3% neutron leakage rate (as reported by Hernandez and Brown 2020). Therefore, in this context, the term “criticality period” refers to the burnup period during which k_inf reaches 1.03 (as denoted by the dashed lines in the figures).

Multiplication factors and criticality periods

WgPuO₂-ThO₂ duplex vs. LEU fuel

A comparison to the LEU fuel is required to investigate the performance of neutronic behavior, discharge burnup, and fuel cycle of the WgPuO₂-ThO₂ duplex fuel when it is incorporated into the VVER assembly. The multiplication factor serves as the fundamental value to evaluate the behavior of these parameters. Fig. 6 depicts the k_inf values of LEU (LEU-Ref) and the duplex fuel design under consideration (Du-Ref) versus burnup. The reactivity of the Du-Ref at the BOC is less than that of LEU-Ref because of the larger thermal neutron absorption cross-section of Th-232 than U-238. Even so, the reactivity of the Du-Ref decreases more slowly with burnup than the reactivity of LEU-Ref due to the formation of fissile U-233, Pu-239 and Pu-241. At the end of the LEU irradiation, the reactivity of the Du-Ref case is still higher than that of LEU-Ref. This implies that the considered duplex fuel has sufficient reactivity to burn for a longer period of time, which gives the duplex fuel an advantage in longer fuel cycle schemes. In terms of criticality periods, the Du-Ref case attains the criticality period at 84.92 GWd/ton, which is equivalent to 5.81 equivalent full power years (EFPYs). The LEU-Ref case, on the other hand, only achieves the criticality period at 36.12 GWd/ton, which corresponds to 2.47 EFPYs.

Figure 6. 

Comparison of k_inf evolution vs. burnup (top), and accumulated burnup over EFPYs (bottom) for reference and duplex assemblies.

Effects of MAs addition on Du-Ref

Next step is to look into the MAs cases trend of k_inf with the reference duplex model. Fig. 7 depicts the results as a function of burnup, where (a) represents the coated cases and (b) represents the mixed cases. Since MA nuclides have large thermal neutron capture cross-sections, the k_inf values at BOC will certainly decrease as the quantity of MAs in the fuel increases along with the displacement of fissile material. Furthermore, the reactivity suppression followed the same trend for both configurations (i.e. coated and mixed), and the suppression increased with the increase of the IFBAs in the assembly. However, as evident in Fig. 8, the decrease rate (|∆k|) determined by equation (1)) was higher in all coated cases, resulting in a marginally significant reactivity suppression when compared to their corresponding values in mixed cases, more precisely, 25% higher on average for all cases. This entire scenario is due to differences in the degree of self-shielding in terms of MAs design and placement, whether homogeneous or heterogeneous. The distribution of MAs over a larger area reduces the self-shielding effect (Evans et al. 2022). As a result, the reactivity suppression in mixed configurations is slightly lower than in coated configurations.

Figure 7. 

k_inf versus fuel burnup for a. Coated cases, and b. Mixed cases.

Figure 8. 

|∆k|, comparison between different loading designs a. Coated cases, and b. Mixed cases.

In terms of the criticality period (cycle length), it follows the same pattern as reactivity suppression, decreasing as the number of MAs and IFBAs in the assembly increases. Yet, another difference between cycle lengths in both configurations is that the cycle length penalty in coated cases is about 1.08% on average, whereas the cycle length penalty in mixed cases is about 1.27% on average. Fig. 9 deficit the criticality periods for the considered cases.

Figure 9. 

Criticality periods, comparison between different loading designs a. Coated cases, and b. Mixed cases.

$\left|\Delta \mathrm{k}\right|=\left|\frac{{\mathrm{k}}_{\mathrm{i}}}{{\mathrm{k}}_{\text{Du \_ Ref }}}-1\right|·100$ (1)

Where k_i is the calculated multiplication factor at BOC after adding MAs, and k_{Du_Ref} is the multiplication factor for the WgPuO₂-ThO₂ duplex case multiplied by 100 to get results with %.

Isotopic concentrations change with burnup

Plutonium concentrations

Fig. 10 illustrates the plutonium isotopic concentrations at BOC and EOC (end of the cycle) for all MAs cases at the same irradiation period. As the Du_Ref case without initial MAs, this case serves as an initial point for further comparison with the other cases.

Figure 10. 

Assembly-averaged plutonium isotopic concentrations at BOC and EOC.

The only isotope of plutonium that exhibits depletion during the simulation is Pu-239 which positively impacts the total depleted plutonium concentrations (Fig. 11). More precisely, the depletion value for Pu-239 in Du _Ref is approximately 88.81% from its initial concentration, and the rate increases as the amount of MAs increases, from 0.02% to 0.32% for coated cases and 0.04% to 0.39% for mixed cases. In terms of Pu-239 weight percent at discharged fuels, Pu-239 in Du _Ref case accounts for approximately 31.74 wt.%, and the addition of MAs reduces this value by 0.13% to 1.38% for coated cases and 0.15% to 1.50% for mixed cases. Otherwise, employing WgPuO₂-ThO₂ fuel increased the concentrations of Pu-238, Pu-240, Pu-241, and Pu-242 by 2002.40%, 143.14%, 1602.14%, and 6834.71%, respectively, throughout burnup. To represent approximately 1.27 wt.%, 42.66 wt.%, 18.03 wt.%, and 6.29 wt.%, respectively, at discharged fuel. The use of MAs increases the accumulation of Pu-238 at discharged fuel by 0.22 wt.% to 2.19 wt.% for coated cases and 0.22 wt.% to 2.20 wt.% for mixed cases, this is primarily due to increased production from Np-237 neutron captures. Pu-242 weight percent at discharged fuel increased by plus 0.01 wt.% to 0.14 wt.% for mixed cases and 0.02 wt.% to 0.15 wt.% for coated cases because Pu-242 is mainly formed from Am-241. Pu-240 and Pu-241 undertook, however, slightly decrease at discharge, by 0.07 wt.% to 0.71 wt.% coated cases and 0.07 wt.% to 0.70 wt.% for mixed cases for Pu-240, and 0.03 wt.% to 0.25 wt.% coated cases and 0.02 wt.% to 0.12 wt.% for mixed cases for Pu-241. Overall, comparing the presented results reveals that the difference between homogeneous and heterogeneous loadings was relatively small.

Figure 11. 

Assembly-averaged total plutonium concentrations at BOC and EOC.

MAs isotopes

The behavior of the analyzed MA isotopes, considered in this study, is shown in Fig. 12. As presented in the graphic representation, only Np-237 concentrations decrease with burnup, whereas Am-241, Am-243, Cm-244, and Cm-245 concentrations accumulate with burnup. Np-237 concentration decreased by 58.30% to 69.58% in coated cases and 58.26% to 69.12% in mixed cases, as a result of its transmutation. The successful transmutations of Np-237 reduce the stockpile because it is the main constituent of the MAs in depleted nuclear fuel. Am-241 accumulation, on the other hand, exhibits very interesting behavior in that its accumulation rate decreases as initial concentration increases, for example, in coated cases the accumulation rate falls from 974.20% to 7.17%, and in mixed cases the accumulation rate falls from 988.59% to 11.89%, from its concentration at BOC. What’s more, when compared to the Du _Ref case, Am-241 concentration at EOC decreased by 0.05% to 0.29% for coated cases and increased by 0.33% to 3.28% for mixed cases, which means that using Am-241 in a coated configuration (heterogeneous) enhance its transmutation. The accumulation rate of the remaining MAs, Am-243, Cm-244, and Cm-245, decreased as initial concentrations increased, similar to Am-241. For Am-243, the accumulation rate decreases from 2369.29% to 164.57% in coated cases and from 2395.28% to 169.51% in mixed cases. For Cm-244, the accumulation rate decreases from 3626.10% to 461.42% in coated cases and from 3600.71% to 423.30% in mixed cases. In the case of Cm-245, the accumulation rate decreases from 8802.17% to 1096.58% in coated cases and from 9244.68% to 1343.07% in mixed cases, from their concentrations at BOC. Unfortunately, at the discharge, their concentrations were higher than those produced by the Du_Ref. Am-243 concentration increased by 0.64% to 7.82% in coated cases and by 0.74% to 8.97% in mixed cases. Cm-244 concentration increased by 6.07% to 59.82% in coated cases and by 4.36% to 47.79% in mixed cases. Cm-245 concentration increased by 4.08% to 39.89% in coated cases and by 8.22% to 67.38% in mixed cases. Besides coated cases produce less Am-243 and Cm-245, but more Cm-244. Based on the results of the simulation calculations, it is strongly advised that future research should focus on the incorporation of enriched minor actinides, specifically NpO₂ or AmO₂. These particular actinides constitute the majority of the total minor actinides found in depleted LWR fuel and display promising transmutation rates.

Figure 12. 

Assembly-averaged MA concentrations at BOC and EOC.

Th-232 and U-233 concentrations

One of the primary goals of using thorium-based fuel is to produce U-233, a valuable fissile material for future use (Colton and Bromley 2018). This is due primarily to the advantageous properties of Th-232, which has a thermal neutron absorption cross-section roughly three times that of U-238. Within the thermal neutron spectrum, this property results in a significantly higher rate of transmutation into fissile U-233. Furthermore, when compared to U-235 and Pu-239, U-233 has the highest neutron reproduction factor. As a result, U-233 has the potential to become a valuable future energy resource, improving the resilience and sustainability of the nuclear fuel cycle while reducing proliferation risks.

The concentrations of Th-232 and U-233 at BOC and EOC for all examined cases are illustrated in Fig. 13. As explained previously in the MAs incorporation design section, MAs loading only directly impacts the WgPuO₂ zone in the duplex cases, resulting in the same initial Th-232 concentration for all cases. The addition of MAs, however, increases U-233 breeding by 0.01% to 0.06% (steadily increasing as MAs loading increases) due to a decrease in fissile Pu-239, which has a high ability to absorb thermal neutrons.

Figure 13. 

Assembly-averaged Th-232 and U-233 concentrations at BOC and EOC.

This study compares the neutronic performance of an LEU fuel and an innovative PuO₂-ThO₂ duplex-based fuel in a VVER-1200 assembly using the DRAGON lattice physics code yielded valuable insights. In this duplex configuration, WgPuO₂ (enriched in Pu-239) was used in the inner perimeter, while pure ThO₂ was used in the outer perimeter, effectively recycling excess Pu-239 and producing U-233. One noteworthy finding was the remarkable increase in burnup, up to 135%, achieved by the duplex fuel when compared to an LEU fuel with an equivalent constant power density. Additionally, the duplex fuel exhibited a reduced reactivity swing throughout the fuel cycle, which is a crucial safety consideration.

The use of MAs to mitigate excess reactivity was also examined, with two MA configurations investigated: one involving coated MAs and the other involving MAs mixed with the WgPuO₂ fuel. Notably, coating MAs with WgPuO₂ fuel resulted in slightly higher reactivity suppression at BOC and a slightly lower burnup penalty when compared to mixing MAs with WgPuO₂ fuel. The study also scrutinized the isotopic composition of the discharged fuel. The use of MAs resulted in significant Pu-239 transmutation, lowering even-numbered plutonium isotopes and increasing resistance to proliferation. However, it also resulted in higher concentrations of Pu-238, Pu-242, Am-243, Cm-244, and Cm-245 in the discharged fuel. Importantly, the reactivity coefficients (FTC, MTC, and BWC) of the investigated cases were mainly negative, with FTC being noticeably more negative in MAs cases due to Doppler broadening from MAs captures. Higher MAs incorporation patterns can also be used to partially substitute or reduce the boron concentration in the moderator because MTCs will be much less negative at higher concentrations.

Future research could also focus on core designs will be explored, with a focus on further validating the feasibility of the fuel and assembly arrangements to enhance the practicality of the duplex fuel. This will encompass investigations into safety parameters and variations in duplex configurations, such as ThO₂-WgPuO₂. Additionally, future research will extend to the examination of alternative burnable absorbers, it will also delve into coupled neutronic and thermal-hydraulic studies to gain a comprehensive understanding of the system’s performance under various conditions.

Achraf Radi: Resources, Data curation, Investigation, Writing - original draft. Ouadie Kabach: Conceptualization, Data curation, Formal analysis, Investigation, Methodology, Supervision, Validation, Writing - review & editing. El Mahjoub Chakir: Supervision, Formal analysis, review & editing.

This paper and the references therein contain all the data to reproduce and validate the presented results.

The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.

The authors highly appreciate the valuable comments and suggestions of the respected unknown reviewers that have improved the quality of this paper