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All issues Volume 11 (2025) EPJ Nuclear Sci. Technol., 11 (2025) 50 Full HTML
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EPJ Nuclear Sci. Technol.
Volume 11, 2025
Special Issue on ‘Overview of recent advances in HPC simulation methods for nuclear applications’, edited by Andrea Zoia, Elie Saikali, Cheikh Diop and Cyrille de Saint Jean
Article Number 50
Number of page(s) 9
DOI https://doi.org/10.1051/epjn/2025043
Published online 03 September 2025

© C. Liu et al., Published by EDP Sciences, 2025

Licence Creative CommonsThis is an Open Access article distributed under the terms of the Creative Commons Attribution License (https://creativecommons.org/licenses/by/4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.

1. Introduction

As one of the six nuclear reactor technologies selected by the Generation IV International Forum (GIF), Lead-cooled Fast Reactor (LFR) has become one of the most promising concepts and drawn extensive attention in nuclear industry. Many LFR designs have been proposed, such as SVBR-100 [1] and BREST-300 [2] in Russia, ALFRED and ELSY [3] in Europe, and SSTAR [4] in the United States.

State Power Investment Corporation Research Institute proposed a Lead-Bismuth Eutectic (LBE) cooled fast reactor BLESS [5] (Breeding Lead-based Economical and Safe System), which envisages to satisfy the public’s demand for a safe, economic and environmental-friendly nuclear power system. In addition to generating nuclear energy, the LBE cooled reactor is also capable of burning nuclear waste through transmutation.

Based on the self-designed LBE cooled fast reactor, the neutronics and thermal hydraulics characteristics were calculated using the Monte-Carlo code cosRMC [6–8] and the subchannel code KMC-SUB [9]. The Monte-Carlo code cosRMC is a part of the COSINE (COre and System INtegrated Engine for design and analysis) [10] code package. The KMC-SUB is a transient sub-channel code for the best estimation of lead cooled reactor cores. In particular, the neutron energy spectrum, neutron flux, transmutation ability, plutonium depletion performance and keff change over the cycle, and the temperature fields of the fuel under steady state were analyzed.

The numerical simulation was conducted in the Milkyway-2. It has approximately 16 000 nodes, and each node has two Xeon E5 series 12 cores CPUs and a Memory of 64 G.

2. Introduction of BLESS-D

The BLESS-D reactor is a pool-type reactor in which the core, main pumps, and steam generators are placed in the reactor vessel, and the entire reactor is filled with liquid LBE. The core is radially divided into three different regions according to different fuel enrichment. Fuel assemblies are surrounded by reflector assemblies. Key parameters of the core design are listed in Table 1.

Table 1.

Main design parameters of BLESS-D.

3. Neutronics characteristics analysis

3.1. Core arrangement

As a fast neutron reactor, it is expected that the BLESS-D’s core have a high conversion ratio and a long core life, which is capable of applying fuel assemblies with a low 235U enrichment, while maintaining sufficient excess reactivity. Therefore, taking the core conversion ratio, the keff of the initial charge and the radial power distribution into consideration for the optimal performance of the core, the enrichment of 235U from the inside to the outside of BLESS-D reactor is designed to be 14%, 16% and 19.75%, respectively. The core arrangement established by cosRMC is shown in Figure 1.

thumbnail Fig. 1.

Core arrangement and overall structure of BLESS-D.

3.2. Change of keff with fuel consumption

The change of keff with burnup at a thermal power of 300 MW is calculated. For a relatively accurate calculation, in particular the burnup change with time is larger in the very beginning; the first burnup step is set to be 20 days, while the succeeding burnup steps are 100 days. The core is radically divided into three burnup zones according to the difference of 235U enrichment, and is axially regarded as one burnup zone.

Figure 2 shows the change of keff with fuel consumption. The initial keff of the core is 1.06711 (0.00013). Without considering the radiation resistance and corrosion resistance of the structural materials of the core, the life of the first cycle is 1770 EFPD, about 5 EFPY, and the average discharge burnup of the core is 50.8 GWd/tU.

thumbnail Fig. 2.

Change of keff with the burnup.

3.3. Core neutron spectrum

The transmutation ability is closely related to the neutron spectrum of the reactor. PWR is a thermal neutron reactor, and the neutron energy is mainly thermal neutron. BLESS-D reactor is a fast neutron spectrum reactor, and is mainly constituted of fast neutrons. As highly dependence of neutron energy, the micro cross sections of fission reaction and radiation capture reaction for each nuclide are different, so the quantities of production and consumption vary. Figure 3 shows the neutron energy spectrums at the beginning of life for BLESS-D and a typical PWR.

thumbnail Fig. 3.

The neutron energy spectrum at the beginning of life for BLESS-D reactor and a typical PWR.

It can be seen that the fast neutron ratio for BLESS-D is relatively larger, while the thermal neutron ratio for PWR is relatively larger. Since the fission cross section under thermal spectrum for 239Pu is larger than that under fast spectrum, as a result, the 239Pu accumulated during burnup for PWR is more probable to fission, compared with that of BLESS-D. This makes the 239Pu mass for PWR smaller than that of BLESS-D.

3.4. Transmutation analysis for UO2

The changes of net mass of 235U, 238U, 239Pu, 240Pu, 237Np, 241Am, 243Am, 243Cm, 244Cm, 245Cm, 137Cs, 99Tc and 129I (net residual mass at certain burnup) are discussed in this section. 237Np, 241Am, 243Am, 243Cm, 244Cm and 245Cm are Minor Actinides (MAs), 137Cs, 99Tc and 129I are Long-Lived Fission Products (LLFPs).

For 239Pu and 240Pu, they were mainly converted from 238U radiation capture reaction. At the same time, 239Pu and 240Pu can undergo fission reaction and radiation capture reaction. The net cumulative amount of 239Pu and 240Pu at the end of life is 2.78 × 105 g and 1.06 × 104 g respectively. From Figure 4, it can be seen that the curve of 239Pu and 240Pu are nearly flat at the end of life, the balance has been reached.

Figure 5 shows the change of LLFP radionuclides. It can be seen that the change of 237Np is much higher than that of other nuclides. At the EOL (end of life), the mass of 237Np is 5.55 × 103 g, while the masses of 241Am, 243Am, 243Cm and 244Cm are much lower, which are 28.12 g, 0.25 g, 1.53 × 10−2 g and 1.44 × 10−2 g, respectively. For LLFP, the masses at the EOL for 137Cs, 99Tc and 129I are 1.96 × 104 g, 1.33 × 104 g and 2.67 × 103 g, respectively, showing the cumulative amounts for LLFP are generally higher than that of MAs.

thumbnail Fig. 4.

Net masses of 235U, 238U, 239Pu and 240Pu with fuel consumption.
thumbnail Fig. 5.

Net accumulation of LLFP and some transuranic elements with fuel consumption.

Table 2 shows the net accumulation the long-lived high-level radionuclides, produced by the heavy metals consumed per unit mass (per ton of heavy metals, denoted tHM) for BLESS-D reactor at an average burnup depth of 33 GWd/tU, and a comparison with that of a PWR loaded with UO2 under identical average burnup depth.

Table 2.

Comparison of major long-lived high-level radionuclides and some transuranic elements produced by BLESS-D and PWR.

Through comparing nuclide accumulations of interest for fast and thermal neutron spectrums, it can be seen that the 239Pu production under fast spectrum in BLESS-D reactor is significantly higher than that of PWR, and BLESS-D performs better in reducing the MAs, while the productions of LLFPs are at the same scale. This presents that BLESS, as a fast reactor, has greater performance in fuel utilization. The main differences in the results are primarily caused by the differences in the neutron spectra of the reactors. As seen from Figure 3, the core of BLESS-D features a larger share of fast neutrons compared with PWR, while the core of PWR has a larger share of thermal neutrons. This implies that BLESS-D operates in the fast neutron region, whereas PWR operates in the thermal neutron region. Due to the capture/fission, cross-sections of minor actinides under fast neutron spectrum are relatively small, and the fission reactions of most minor actinides are high-threshold reactions, this make the fast neutron spectrum more conducive to the fission of minor actinides, resulting in a smaller production of minor actinides in BLESS-D compared with PWR. The loss of long-lived fission products in BLESS-D results entirely from capture reactions, and a softer neutron spectrum can increase the capture cross-section of long-lived fission products. Thus, compared with thermal neutron reactors, fast neutron reactors do not have an advantage in reducing long-lived fission products.

3.5. Core arrangement for MOX

MOX fuel can be manufactured from plutonium recovered from PWR spent fuel, mixed with depleted uranium. The efficiency of the use of the spent fuel unloaded from PWRs and storage of MAs with long half-lives have a great impact on the sustainability of nuclear energy. Therefore, it is important to conduct a deep study on the features of fast reactors loaded with MOX fuel.

In the present section, the characteristics of BLESS-D loaded with MOX are investigated.

The layout of MOX(UO2+PuO2) core loading is the same as the UO2 core loading as shown in Figure 1. The mass ratios of PuO2 from the inside to the outside of BLESS-D reactor are 16%, 18% and 22%, respectively. MAs are also added to undergo transmutations from the interaction of the fast neutrons and MAs.

Figure 6 shows the change of keff with fuel consumption. The initial keff of the core is 1.04512 (0.00013). Without considering the radiation resistance and corrosion resistance of the structural materials in the core, the life is 3000 EFPD, about 8.2 EFPY, and the average discharge burnup of the core is 84.8 GWd/tU.

thumbnail Fig. 6.

Net change of keff with burnup.

Comparing the EFPD values, while keff is close to 1.0, it could be found that the EFPD value loaded with MOX fuel is about 66.7% larger than that of UO2, showing a great performance of the neutronics for MOX fuel.

3.6. Reactivity coefficient

Table 3 gives the results of reactivity coefficients for UO2 and MOX, at the beginning of life (BOL), the middle of life (MOL), and the end of life (EOL). From the results, it can be concluded that at the beginning, in the middle and at the end of life of the core, the fuel Doppler coefficient, the fuel temperature coefficient, the coolant temperature coefficient and the coolant voiding coefficient are all negative, satisfying the design criteria.

Table 3.

Reactivity coefficients and kinetic parameters.

However, compared with the core loaded with UO2, the safety performance of the MOX core is decreased, mainly due to the decreased loading ratio of 238U, resulting in the decreased resonance self-shielding effect of the core in the superheated neutron spectrum. Therefore, optimizing the fuel-loading scheme, in particular balancing the fuel’s safety performance and the loading quantity of MAs, would be part of future work.

3.7. Transmutation analysis for MOX

The net masses of some key nuclides including 234U, 235U, 236U, 238U, 238Pu, 239Pu, 240Pu, 241Pu and 242Pu in the core with burnup are also analyzed; these nuclides are mainly used to perform fission reactions. Figure 7 shows the changes of net masses of 234U, 235U, 236U and 238U in the core with burnup, and Table 4 summarizes the masses of 234U, 235U, 236U and 238U in the core at the BOL and EOL. It can be seen that the masses of 234U and 236U in the whole cycle life increase by 6.04 × 103 g and 3.70 × 103 g respectively, 25.0 times and 4.6 times than those at the BOL. During the whole cycle life, the mass consumptions of 235U and 238U are 1.61 × 104 g and 7.14 × 105 g respectively, which are 49.3% and 8.9% lower than those at the BOL. Figure 8 shows the changes of the net mass of 238Pu, 239Pu, 240Pu, 241Pu and 242Pu in the core with burnup, and Table 5 shows the masses of 238Pu, 239Pu, 240Pu, 241Pu and 242Pu in the core at the BOL and the EOL.

Table 4.

Comparison of masses of 234U, 235U, 236U and 238U at the BOL and EOL for MOX.

Table 5.

Comparison of masses of 238Pu, 239Pu, 240Pu, 241Pu and 242Pu at the BOL and EOL for MOX.

thumbnail Fig. 7.

Net mass changes of 234U, 235U, 236Uand 238U with burnup for MOX.
thumbnail Fig. 8.

Net mass changes of 238Pu, 239Pu, 240Pu, 241Pu and 242Pu with burnup for MOX.

Table 6.

Comparison of masses of 237Np, 241Am, 243Am, 243Cm, 244Cm and 245Cm at the BOL and EOL for MOX.

It can be seen from Figure 8 and Table 5 that the mass of 239Pu gradually decreases with burnup, the masses of 240Pu and 242Pu remain almost constant, reaching an equilibrium state.

The quantity of 238Pu increases, mainly due to the radiation capture reactions of 237Np, 241Am, 242Cm and other MAs, making the production of 238Pu greater than its consumption.

In addition, the mass of 241Pu decreases slightly with burnup. 241Pu can be produced by the radiation capture reaction of 240Pu, and 241Pu can be consumed into 242Pu through the radiation capture reaction at the same time. As the burnup increases, the reaction chain reaches equilibrium, 241Pu transforms into 241Am through β decay, with a half-life of 14.7 years.

The net masses of 237Np, 241Am, 243Am, 243Cm, 244Cm and 245Cm with burnup are also studied. Figure 9 shows the changes of the net masses of six MAs with burnup, and Table 6 shows the masses of 237Np, 241Am, 243Am, 243Cm, 244Cm and 245Cm in the core at BOL and EOL.

thumbnail Fig. 9.

Net masses changes of 237Np, 241Am, 243Am, 243Cm, 244Cm and 245Cm with burnup for MOX.

It can be seen from Figure 9 and Table 6 that the quantities of 237Np, 241Am and 243Am gradually decrease, and the values at the EOL are 58%, 71% and 82% of those at the BOL, respectively. The quantities of 243Cm, 244Cm and 245Cm increase gradually, and the quantity at the EOL are 315%, 120% and 336% of those at the BOL, respectively.

The total mass of MAs added at the BOL is 6.02 × 105 g, and the total mass of MAs at the EOL is 4.11 × 105 g, that is 68.27% of the total mass at the BOL, showing that the total mass of MAs decreases, demonstrating the transmutation capability of BLESS-D.

To quantitatively assess the transmutation capability of the BLESS reactor core, it is essential to define some indicators for evaluating the transmutation effect of MAs. The transmutation rate refers to the ratio of the mass of MAs transmuted during a cycle to the initial MAs loading at the beginning of the cycle, representing the transmutation fraction. The specific consumption denotes the ratio of the mass of MAs transmuted during the cycle to the total power output of the reactor core, indicating the transmutation rate. The support ratio is the ratio of the transmutation rate of MAs in the cycle to the MAs production rate for a thermal reactor with an equivalent power, representing the transmutation demand. In the present study, it is assumed that for a pressurized water reactor the MAs production rate is 20 kg⋅GWe−1⋅a−1 [11], and the thermal efficiency is 35%, this results in the MAs production rate as 7 kg⋅GWt−1⋅a−1. As computation, the transmutation rate, specific consumption, and support ratio are 31.7%, 77.4 kg⋅GWt−1⋅a−1, and 11.1, respectively. This indicates that adding a small amount of MAs to the BLESS-D reactor core can significantly improve the transmutation performance.

4. Thermal hydraulics characteristics

In this section, the fuel assembly with the largest thermal power and the subchannel with the largest outlet temperature in this fuel assembly are investigated.

As investigated, under normal conditions, the flow velocity of design for the coolant is recommended to be less than 2 m/s [12], and further set to be less than 1.5 m/s for conservative reasons. The cladding temperature of design is recommended to be less than 550° under normal conditions [12] and 650° under certain transient conditions. As for the pellet temperature, it is recommended to be less than 2164° [13], since the melting temperature of fresh UO2 is about 2804°, and decreases by 32° per 10 000 MWd/tU as burnup.

The thermal-hydraulic performance of the fuel assembly with the largest thermal power in the core, which has a power factor of 1.3, is studied using a subchannel code.

In the subchannel calculation, the fuel assembly is divided into 258 subchannels and 127 fuel pins. Figure 10 shows its division and numbering scheme. The power rate of each pin is shown in Figure 11. It shows that the maximum peak power rate is 1.023, and is located near the center of the fuel assembly.

thumbnail Fig. 10.

Subchannel division and numbering of the fuel assemblies.
thumbnail Fig. 11.

Radial power rates distribution of the assembly.

As for the input for calculation, the mass flow rate is equally distributed for each fuel pin. The inlet temperature of 340° is set as the boundary condition.

The outlet temperature in each subchannel is shown in Figure 12. Through calculations, it has been found that the core average outlet temperature is 481.04°, which is very close to the designed value of 490°, with a difference of about 2.05%. It also shows that the power distribution of the outlet temperature distribution presents a rather symmetric shape, and the fuel pins close to the center have a larger value, while the fuel pins in the periphery is relatively small. This is mainly attributed to the difference in power/flow ratio for each channel, lower at the periphery, and reinforced by the cold wall effect. The largest difference in outlet temperature between adjacent channels is 23.51°, which is between subchannel 232 and subchannel 246. This temperature difference is very small, showing a satisfactory thermal performance of the core’s design.

thumbnail Fig. 12.

Core outlet temperature distribution in °C.

For the assembly with the largest power, the coolant temperature distribution in axial direction in 14 subchannels is shown in Figure 13. We can see that the coolant temperature of each channel increases gradually with the flow, and the outer temperatures span from 500° to 550°, satisfying relevant design criterion.

thumbnail Fig. 13.

Axial temperature distribution in 15 different subchannels.

The maximum temperature of the coolant is observed for subchannel 159, near the hottest rod Rod78, which is 548.43°. The axial temperature distributions of the coolant in subchannel 159 are shown in Figure 14. The maximum temperature of the fuel pellet for the subchannel 159 is 1016.34°, satisfying the design criterion under normal operating conditions. The maximum coolant flow velocity in the core is 1.14 m/s, less than the 1.5 m/s recommended value.

thumbnail Fig. 14.

Axial temperature distribution in the subchannel 159.

The maximum cladding temperature is 555.81°, 1.06% higher than the recommended value under normal operating condition, but still well below 650° under certain transient conditions, which is also acceptable as this channel is the hottest in the whole core.

5. Conclusion

The Generation 4th nuclear power system is an important solution for the nuclear energy development in the future, and the LFR is one of the most promising concepts. This paper presents the design of a 300 MW LBE-cooled fast reactor; the core is modelled by cosRMC, a Monte Carlo code.

  • (1)

    The results show that the fuel Doppler coefficient and the coolant-voiding coefficient are all negative, the coolant density change has a small effect on the reactivity; proving good inherent neutronics and safety performance of BLESS-D.

  • (2)

    Compared with PWR, the 239Pu production for BLESS-D is significantly higher, and BLESS-D performs better in reducing the MAs, while the productions of LLFPs are at the same scale.

  • (3)

    The power distribution is rather flat; the peak power factor in the radial direction is 1.3, coolant temperature in the fuel assembly with the largest thermal power, the pellet and cladding temperature in the subchannel with the largest outlet temperature, are all under recommended values, presenting a satisfactory feature of the core’s design.

  • (4)

    The active fuel zone of BLESS-D using MOX features a relatively harder neutron spectrum. When adding a small amount of MA (for example, only adding a share of 5%), the core will have a good performance in transmuting MAs.

Acknowledgments

COSINE is a registered trademark of SNPTC. The authors want to thank all the funding sources, which support the research and development of cosRMC during these years.

Funding

The authors thank SPIC for partial financial support.

Conflicts of interest

The authors declare that they have no competing interests to report.

Data availability statement

This article has not generated nor analyzed data associated with this article.

Author contribution statement

Chunyuan Liu: Methodology, Original Draft Writing, Review, Editing; Xiaosong Chen: Simulations and Original Draft Writing, Review, Supervision; Eing Yee Yeoh: Simulations, Review and Editing; Linsen Li: Project Administration; Yueting Xiao: Simulations, Visualization, Original Draft Writing.

References

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Cite this article as: Chunyuan Liu, Xiaosong Chen, Eing Yee Yeoh, Linsen Li, Yueting Xiao, Neutronics and thermal hydraulics analysis on a Lead-bismuth-cooled fast reactor, EPJ Nuclear Sci. Technol. 11, 50 (2025). https://doi.org/10.1051/epjn/2025043.

All Tables

Table 1.

Main design parameters of BLESS-D.

In the text
Table 2.

Comparison of major long-lived high-level radionuclides and some transuranic elements produced by BLESS-D and PWR.

In the text
Table 3.

Reactivity coefficients and kinetic parameters.

In the text
Table 4.

Comparison of masses of 234U, 235U, 236U and 238U at the BOL and EOL for MOX.

In the text
Table 5.

Comparison of masses of 238Pu, 239Pu, 240Pu, 241Pu and 242Pu at the BOL and EOL for MOX.

In the text
Table 6.

Comparison of masses of 237Np, 241Am, 243Am, 243Cm, 244Cm and 245Cm at the BOL and EOL for MOX.

In the text

All Figures

thumbnail Fig. 1.

Core arrangement and overall structure of BLESS-D.
In the text
thumbnail Fig. 2.

Change of keff with the burnup.
In the text
thumbnail Fig. 3.

The neutron energy spectrum at the beginning of life for BLESS-D reactor and a typical PWR.
In the text
thumbnail Fig. 4.

Net masses of 235U, 238U, 239Pu and 240Pu with fuel consumption.
In the text
thumbnail Fig. 5.

Net accumulation of LLFP and some transuranic elements with fuel consumption.
In the text
thumbnail Fig. 6.

Net change of keff with burnup.
In the text
thumbnail Fig. 7.

Net mass changes of 234U, 235U, 236Uand 238U with burnup for MOX.
In the text
thumbnail Fig. 8.

Net mass changes of 238Pu, 239Pu, 240Pu, 241Pu and 242Pu with burnup for MOX.
In the text
thumbnail Fig. 9.

Net masses changes of 237Np, 241Am, 243Am, 243Cm, 244Cm and 245Cm with burnup for MOX.
In the text
thumbnail Fig. 10.

Subchannel division and numbering of the fuel assemblies.
In the text
thumbnail Fig. 11.

Radial power rates distribution of the assembly.
In the text
thumbnail Fig. 12.

Core outlet temperature distribution in °C.
In the text
thumbnail Fig. 13.

Axial temperature distribution in 15 different subchannels.
In the text
thumbnail Fig. 14.

Axial temperature distribution in the subchannel 159.
In the text

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