Motivated by the growth of global energy demand and the goal of carbon neutrality, lead-cooled fast reactors, which are core reactor types of fourth-generation nuclear energy systems, have become a global research hotspot due to their advantages of high safety, nuclear fuel breeding capability, and economic efficiency. However, its engineering implementation faces key challenges, such as material compatibility, closed fuel cycles, and irradiation performance of structures. This paper comprehensively reviews the latest progress in the core design of lead-cooled fast reactors in terms of the innovation of nuclear fuel, optimization of coolant, material adaptability, and design of assemblies and core structures. The research findings indicate remarkable innovation trends in the field of lead-cooled fast reactor core design, including optimizing the utilization efficiency of nuclear fuel based on the nitride fuel system and the traveling wave burnup theory, effectively suppressing the corrosion effect of liquid metal through surface modification technology and the development of ceramic matrix composites; replacing the lead-bismuth eutectic system with pure lead coolant to enhance economic efficiency and safety; and significantly enhancing the neutron economy and system integration degree by combining the collaborative design strategy of the open-type assembly structure and control drums. In the future, efforts should be made to overcome the radiation resistance of materials and liquid metal corrosion technology, develop closed fuel cycle systems, and accelerate the commercialization process through international standardization cooperation to provide sustainable clean energy solutions for basic load power supply, high-temperature hydrogen production, ship propulsion, and other fields.

This study presents an innovative solution method for ultra-fine group slowing-down equations tailored to stochastic media with double heterogeneity (DH), focusing on advanced nuclear fuels such as fully ceramic microencapsulated (FCM) fuel and Mixed Oxide (MOX) fuel. Addressing the limitations of conventional resonance calculation methods in handling DH effects, the proposed UFGSP method (the ultra-fine group slowing-down method with the Sanchez–Pomraning method) integrates the Sanchez–Pomraning technique with the ultra-fine group transport theory to resolve spatially dependent resonance cross-sections in both matrix and particle phases. The method employs high-fidelity geometric modeling, iterative cross-section homogenization, and flux reconstruction to capture neutron self-shielding effects in stochastically distributed media. Validation across seven FCM fuel cases, four poison particle configurations (BISO/QUADRISO, Bi/Tri-structural Isotropic), and four plutonium spot problems demonstrated exceptional accuracy, with maximum deviations in effective multiplication factor k_eff and resonance cross-sections remaining within ±138 pcm and ±2.4%, respectively. Key innovations include the ability to resolve radial flux distributions within TRISO particles and address resonance interference in MOX fuel matrices. The results confirm that the UFGSP method significantly enhances computational precision for DH problems, offering a robust tool for next-generation reactor design and safety analysis.

This study presents an optimized subgroup transport equation solving method to enhance the computational efficiency of resonance calculations in high-fidelity reactor core simulations. By consolidating all the resonance groups into an equivalent single group and performing fixed-source calculations only for the representative subgroup cross-sections, this method significantly reduced the computational burden compared to the conventional subgroup method. Validation studies on single-cell, 2D assembly, and 3D assembly problems demonstrated that the proposed method achieves computational accuracy comparable to the conventional approach while requiring fewer fixed-source equations. This advancement offers a promising solution for improving the efficiency of resonance calculations in high-fidelity reactor core simulations, paving the way for more accurate and computationally efficient modeling of complex reactor systems.

This paper addresses the ill-posed problem in calculating subgroup parameters for resonance self-shielding within nuclear reactor physics. The conventional Pade approximation method often yields negative subgroup cross-sections lacking physical meaning due to its treatment of overdetermined nonlinear systems, making the subgroup transport equations unsolvable. To overcome this, an optimized Pade approximation method is proposed: a resonance factor criterion is used to select energy groups requiring calculation; a systematic procedure dynamically traverses background cross-section combinations starting from a minimal subgroup number, incrementally increasing it until solutions meeting accuracy constraints with positive parameters are found; and, given the insufficiency of background points, a high-resolution resonance integral table is constructed, particularly for ranges exhibiting significant cross-section variations. Numerical validation confirms the method eliminates negative parameters, ensures physical validity, and significantly improves accuracy across benchmark cases including typical fuel pins, burnt pellets, and Gd-bearing lattices. This approach effectively resolves the ill-posedness of the traditional method, offering a more robust and precise subgroup resonance treatment for high-fidelity core neutronics.

Small nuclear power reactors have small core dimensions, frequent power changes, and more severe power distortion compared to nuclear power stations. However, their core has fewer measurement points, making it difficult to observe their core power distribution. High-precision physical calculation programs can accurately calculate the core power distribution, but the real-time performance of the calculation is poor, which is not conducive to online use. In this study, based on physical computing programs, the power distribution spectrum library of small nuclear power reactors under different operating conditions is calculated, and artificial intelligence algorithms are designed. A data-driven model for the proxy relationship between operating state parameters and core power distribution is trained and constructed to achieve rapid calculation and online support of core power distribution, which improves the level of online safety supervision of small power reactors. Numerical experiments show that this method has high accuracy and good robustness, and can meet the requirements of small nuclear power reactor operation safety support. This research is based on a data-driven proxy model and has achieved fast computation of power distribution in the fuel cores of small modular reactors. It addresses the issue of insufficient real-time performance of high-precision physical programs and has important significance for the safe operation of reactors.

The design of a long-life reactor must not only achieve high burnup and closed fuel cycle utilization but also account for system heat transfer efficiency and inherent safety characteristics, thereby enhancing the long-term operational reliability of the reactor. To address these requirements, this paper proposes a long-life core design scheme based on spent-fuel plutonium-thorium nitride fuel and lead-bismuth eutectic coolant. This scheme exploits the excellent breeding and burnup performance of thorium-based fuel under a fast neutron spectrum and employs the neutronics calculation code DRAGON and the subchannel code COBRA-PB to carry out corresponding core physics analysis and safety characteristic assessment. The results show that the core can achieve a 25-year long refueling cycle, with mild reactivity swing over the entire lifetime and low control difficulty; the core radial power gradually shifts inward with increasing burnup, achieving power flattening and thereby mitigating safety risks caused by localized deep burnup; both the control assembly and shutdown assembly possess sufficient reactivity margins, with a single set of control mechanisms capable of achieving effective shutdown upon independent insertion; all key reactivity coefficients of the core are negative, demonstrating negative-feedback regulation capability and inherent safety characteristics. Calculations indicate that the cladding temperatures of both the average channel and the hottest channel remain within limits, leaving ample safety margins, ensuring long-term cladding mechanical integrity, and effectively preventing high-temperature nitride fuel from melting through the cladding, resulting in outstanding overall core safety performance. This core design scheme can provide a theoretical basis and design reference for the engineering application of long-life reactors. It should be noted that this study is an exploratory investigation at the conceptual design stage of the core. The analysis is conducted on the basis that the fuel supply remains stable, that steady-state calculation models are adopted, and that the performance data of key materials are taken from published literature. Accordingly, the conclusions drawn still require further support from experimental verification and detailed design. Nevertheless, the proposed core design can provide a theoretical basis and design reference for the engineering application of long-life reactors.