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Advanced Nuclear Energy Systems: Design, Analysis and Calculation Methods

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A special issue of Energies (ISSN 1996-1073). This special issue belongs to the section "B4: Nuclear Energy".

Deadline for manuscript submissions: 20 July 2026 | Viewed by 2532

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Special Issue Editors

Dr. Qian Zhang

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Guest Editor

Laboratory for Advanced Nuclear Energy Theory and Applications, Zhejiang Institute of Modern Physics, Department of Physics, Zhejiang University, Hangzhou 310027, China
Interests: nuclear energy systems; nuclear reactor optimization design; high-resolution numerical reactor technology

Dr. Yunfei Zhang

E-Mail Website
Guest Editor

Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China
Interests: reactor physics methods; reactor design; fuel cycle and burnup calculation methods; radioactive source term; reactor physics calculation software

Special Issue Information

Dear Colleagues,

As the world accelerates its transition toward a low-carbon future, advanced nuclear energy systems, including Generation IV reactors, Small Modular Reactors (SMRs), and fusion concepts, are poised to play a pivotal role. The successful development and deployment of these systems hinge on groundbreaking innovations in reactor design, the accuracy of safety analyses, and the sophistication of computational and calculation methods. These next-generation reactors present unique challenges, from complex multi-physics phenomena and advanced materials to enhanced safety requirements and economic viability.

This Special Issue aims to gather high-quality, original research that addresses the key challenges in the lifecycle of advanced nuclear energy systems. We invite contributions on a wide range of topics, including but not limited to innovative reactor core and system designs, advanced thermal–hydraulic and neutronic analysis, high-fidelity modeling and simulation, verification and validation of computational tools, uncertainty quantification, and the application of artificial intelligence and machine learning in nuclear engineering. We hope to provide a premier platform on which researchers may share their latest findings, fostering collaboration and driving the future of clean and sustainable nuclear energy.

Dr. Qian Zhang
Dr. Yunfei Zhang
Guest Editors

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Keywords

advanced nuclear reactors
generation IV reactors
small modular reactors (SMRs)
reactor design and safety analysis
neutronics and reactor physics
thermal–hydraulics
advanced calculation methods
high-fidelity simulation
uncertainty quantification
artificial intelligence in nuclear engineering

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Published Papers (4 papers)

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Research

20 pages, 2244 KB

Open AccessArticle

Critical Benchmark Validation of the Core Physics Multigroup Cross-Section Library TPEX

by Ying Chen, Haicheng Wu, Lili Wen, Yue Xiao, Jinchao Zhang, Qian Zhang, Xiaofei Wu and Huanyu Zhang

Energies 2026, 19(9), 2143; https://doi.org/10.3390/en19092143 - 29 Apr 2026

Viewed by 188

Abstract

Core physics multigroup cross-section libraries provide essential cross-section and burnup data for reactor neutron physics calculations, serving as a fundamental prerequisite for reactor physics analysis. The China Nuclear Data Center has developed the TPEX multigroup cross-section library for pressurized water reactors (PWRs) based on the Chinese Evaluated Nuclear Data Library CENDL-3.2. A systematic critical benchmark validation of the newly developed TPEX library has been performed. To verify its applicability and accuracy, the validation has been conducted against 131 critical benchmark experiments from the International Criticality Safety Benchmark Evaluation Project (ICSBEP 2006) and the WIMS-D library update project. The calculated effective multiplication factors

(k_{e f f})

are compared with the experimental values, results from equivalent multigroup libraries, and reference solutions from Monte Carlo code. The results indicate that the absolute average deviations between the calculated

k_{e f f}

values using the TPEX library and the experimental measurements are 280 pcm for the uranium solution experiments, 410 pcm for the plutonium solution experiments, 10 pcm for the uranium metal lattice experiments, 20 pcm for the uranium dioxide lattice experiments, 22 pcm for the MOX fuel lattice experiments, and 150 pcm for the LCT001 uranium oxide assembly experiments. Accordingly, the TPEX library demonstrates excellent performance in reactivity predictions for PWRs. Full article

(This article belongs to the Special Issue Advanced Nuclear Energy Systems: Design, Analysis and Calculation Methods)

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22 pages, 3097 KB

Open AccessArticle

Preliminary Neutronic Design and Thermal-Hydraulic Feasibility Analysis for a Liquid-Solid Space Reactor Using Cross-Shaped Spiral Fuel

by Zhichao Qiu, Kun Zhuang, Xiaoyu Wang, Yong Gao, Yun Cao, Daping Liu, Jingen Chen and Sipeng Wang

Energies 2026, 19(7), 1811; https://doi.org/10.3390/en19071811 - 7 Apr 2026

Viewed by 623

Abstract

As the key technology of space exploration, space power has been a major area of international research focus. A lot of research work has been carried out around the world for the space nuclear reactor using the heat pipe, liquid metal and gas cooling methods. With the development of molten salt reactor in the Generation IV reactor system, molten salt dissolving fissile material and acting as a coolant at the same time has become a new cooling scheme, which provides new ideas for the design of space nuclear reactors. In this study, a novel reactor, the liquid-solid dual-fuel space nuclear reactor (LSSNR) was preliminarily proposed, combining the molten salt fuel and cross-shaped spiral solid fuel to achieve the design goals of 30-year lifetime and an active core weight of less than 200 kg. Monte Carlo neutron transport code OpenMC based on ENDF/B-VII.1 library was employed for neutronics design in the aspect of fuel type, cladding material, reflector material and the spectral shift absorber. Then, the thickness of the control drum absorber was optimized to meet the requirement of the sufficient shutdown margin, lower solid fuel enrichment, and 30-effective-full power-years (EFPY) operation lifetime. Finally, UC solid fuel with U-235 enrichment of 80.98 wt.% and B₄C thickness of 0.75 cm were adopted in LSSNR, and BeO was adopted as the reflector and the matrix material of the control drum. A spectral shift absorber Gd₂O₃ was used to avoid the subcritical LSSNR returning to criticality in a launch accident. The k_eff with the control drum in the innermost position is 0.954949, and the k_eff reaches 1.00592 after 30 EFPY of operation. The total mass of the active core is 158.11 kg. In addition, the thermal-hydraulic feasibility of LSSNR using cross-shaped spiral fuel was analyzed based on a 4/61 reactor core model. The structure of cross-shaped spiral fuel achieves enhanced heat transfer by generating turbulence, which leads to a uniform temperature distribution of the coolant flow field and reduces local temperature peaks. Based on the LSSNR scheme, some neutronic characteristics were analyzed. Results demonstrate that the LSSNR has strongly negative reactivity coefficients due to the thermal expansion of liquid fuel, and the fission gas-induced pressure meets safety requirements. One hundred years after the end of core life, the total radioactivity of reactor core is reduced by 99% and is 7.1305 Ci. Full article

(This article belongs to the Special Issue Advanced Nuclear Energy Systems: Design, Analysis and Calculation Methods)

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32 pages, 2415 KB

Open AccessArticle

Compilation of a Prediction-Based Validation Dataset for Heat Transfer Modeling of the Paks Spent Fuel Interim Storage Facility

by Attila Érchegyi and Ervin Rácz

Energies 2026, 19(5), 1124; https://doi.org/10.3390/en19051124 - 24 Feb 2026

Viewed by 480

Abstract

This study presents and systematizes a high-reliability measurement and technological dataset suitable for prediction-based validation of the Spent Fuel Interim Storage Facility (SFISF) of the Paks Nuclear Power Plant. The primary objective of this dataset is not the validation of a general-purpose software tool, but to establish a reproducible experimental basis for the objective and quantitative validation of a three-dimensional, facility-scale heat transfer and buoyancy-driven flow model of the SFISF, developed using the finite difference method (FDM), in a passively cooled system where heat conduction, thermal radiation, and natural convection simultaneously occur. The applied measurement systems (SMAS, CTRS, and the in-house developed CFEPR), their spatial arrangement, accuracy characteristics, as well as data post-processing and the generation of model execution inputs are described in detail. Special emphasis is placed on the functional separation of the available data into initialization data, model execution data, and independent validation datasets, ensuring that model assessment does not rely on calibration or parameter fitting. Furthermore, the estimation of decay heat generated by the stored fuel assemblies is presented using both a standard correlation method (ANSI/ANS-5.1) and isotope inventory-based calculations, and the discrepancies between these approaches are treated as input uncertainties and sensitivity analysis factors. The spectral solar load is considered based on the ASTM G-173 reference spectrum, while during cloudy periods an effective irradiance estimation derived from on-site lux measurements is applied. The results indicate that the available measurement and technological information is sufficient for supporting reproducible, transparent, and quantitative validation studies of the three-dimensional numerical model of the SFISF, as well as for assessing the impact of dominant input uncertainties. Full article

(This article belongs to the Special Issue Advanced Nuclear Energy Systems: Design, Analysis and Calculation Methods)

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18 pages, 2961 KB

Open AccessArticle

Experimental Design and Numerical Analysis of Volume Internal Heat Generation Source in Fluids Based on Microwave Heating

by Shanwu Wang, Hui Deng, Jian Tian, Pinyan Huang, Hongxiang Yu, Shuaiyu Xue, Ying Cao, Chong Zhou and Yang Zou

Energies 2026, 19(1), 172; https://doi.org/10.3390/en19010172 - 28 Dec 2025

Viewed by 650

Abstract

Liquid-fueled molten salt reactors (MSRs) are characterized by the use of liquid nuclear fuel, which leads to a unique thermal-hydraulic phenomenon in the core involving the simultaneous occurrence of nuclear fission heat generation and convective heat transfer. This distinctive behavior creates a critical need for high-fidelity experimental data on internally heated flows, yet such studies are severely constrained by the lack of methods to generate controllable, high-power-density volumetric heat sources in fluids. To address this methodological gap, this study proposes and numerically investigates a novel experimental concept based on microwave heating. The design features an innovative multi-tier hexagonal resonant cavity with fifteen strategically staggered magnetrons. A comprehensive multi-physics model was developed using COMSOL Multiphysics to simulate the coupled electromagnetic, thermal, and fluid flow processes. Simulation results confirm the feasibility of generating a volumetric heat source, achieving an average power density of 6.9 MW/m³. However, the inherent non-uniformity in microwave power deposition was quantitatively characterized, revealing a high coefficient of variation (COV) for power density. Crucially, parametric studies demonstrate that this non-uniformity can be effectively mitigated by optimizing the flow channel geometry. Specifically, using a smaller diameter tube or an annulus pipe significantly improved temperature field uniformity, reducing the temperature COV by over an order of magnitude, albeit at the cost of reduced absorption efficiency. Preliminary discussion also addresses the extension of this approach towards molten salt experiments. The findings establish a practical design framework and provide quantitative guidance for subsequent experimental investigations into the thermal-hydraulic behavior of internally heated fluids, offering a promising pathway to support the design and safety analysis of liquid-fueled MSRs. Full article

(This article belongs to the Special Issue Advanced Nuclear Energy Systems: Design, Analysis and Calculation Methods)

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