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Advanced Technologies in Nuclear Engineering

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

Deadline for manuscript submissions: 31 December 2024 | Viewed by 8049

Special Issue Editors

Prof. Dr. Guglielmo Lomonaco

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Guest Editor
Department of Mechanical, Energy, Management and Transport Engineering (DIME), University of Genoa, Via all'Opera Pia, 15A, 16145 Genova, GE, Italy
Interests: nuclear energy; nuclear technology; innovative nuclear fuel cycles; neutronics; CFD; advanced nuclear systems; energy scenarios; nuclear hydrogen production; HTR; LFR; GFR; ADS; SMR; nuclear space reactors
Special Issues, Collections and Topics in MDPI journals
Prof. Dr. Álvaro Rodríguez-Prieto

E-Mail Website
Guest Editor
Department of Manufacturing Engineering, Universidad Nacional de Educación a Distancia, Juan del Rosal 12, E28040 Madrid, Spain
Interests: materials processing technologies; metal forming; additive manufacturing; materials technology; data-driven decision methodologies; materials selection in manufacturing; equipment reliability; failure prognosis; nuclear power applications; renewable energy applications; oil & gas applications; aerospace applications;industrial heritage
Special Issues, Collections and Topics in MDPI journals
Dr. Fabio Panza

E-Mail Website
Guest Editor
Department of Fusion and Nuclear Safety Technology, Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), 00123 Rome, Italy
Interests: neutronics (for fission and fusion); diagnostics and fusion fission hybrid systems; Neutronics and nuclear inventory codes; nuclear measurements; nuclear transmutation and tritium breeding

Special Issue Information

Dear Colleagues,

Advanced Technologies in Nuclear Engineering have revolutionized the field of nuclear power generation and have greatly contributed to the advancements in energy production, safety, and waste management. Through continuous research and development, scientists and engineers have been able to harness the power of nuclear energy in a more efficient, sustainable, and secure manner.

Furthermore, advanced nuclear technologies have contributed to the optimization of nuclear fuel cycles. For example, the development of advanced fuel materials, such as mixed oxide (MOX) and metallic fuels, has improved fuel performance and increased the overall efficiency of nuclear reactors. Additionally, advancements in reprocessing techniques, such as pyroprocessing and advanced solvent extraction methods, have facilitated the recycling of spent nuclear fuel, reducing the volume of high-level waste and maximizing the utilization of valuable resources.

This Special Issue aims to address the role of nuclear energy in a future net-zero electricity market that may feature a high presence of renewables and other variable sources and inexpensive peaking capacity. It will cover nuclear power plant design and operation, as well as related technologies. It will provide a forum to discuss and present recent research results, technologies, and best practices on nuclear power plants and their most relevant equipment and components for both fission and fusion technologies, as well as consider their future developments. Papers can include small modular reactor designs, technologies, and operations. Research results on advanced and innovative nuclear fuel cycles will also be included. This Special Issue will also give particular attention to fusion–fission hybrid reactor technologies because these systems could represent an interesting synthesis between nuclear technologies.

Finally, this Special Issue will also bridge research with educational programs, as well as engineering practices, in all disciplines related to nuclear technology.

Prof. Dr. Guglielmo Lomonaco
Prof. Dr. Álvaro Rodríguez-Prieto
Dr. Fabio Panza
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Energies is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2600 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • advanced nuclear systems (including space nuclear reactors)
  • nuclear fuel cycles
  • nuclear fusion and fission technologies (including fusion-fission hybrid reactors)
  • nuclear power plants (including SMR) design and operation
  • nuclear power plants economics
  • nuclear power plants integration with the grid
  • nuclear reactor engineering
  • nuclear reactor physics
  • nuclear safety and security
  • radiation detection and protection systems
  • sustainability of nuclear energy systems

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Published Papers (12 papers)

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12 pages, 6218 KiB  
Article
Multiscale Modeling of Silicon Carbide Cladding for Nuclear Applications: Thermal Performance Modeling
by Gyanender Singh, Jianguo Yu, Fei Xu, Tiankai Yao and Peng Xu
Energies 2024, 17(23), 6124; https://doi.org/10.3390/en17236124 - 5 Dec 2024
Viewed by 94
Abstract
The complex multiscale and anisotropic nature of silicon carbide (SiC) ceramic matrix composite (CMC) makes it difficult to accurately model its performance in nuclear applications. The existing models for nuclear grade composite SiC do not account for the microstructural features and how these [...] Read more.
The complex multiscale and anisotropic nature of silicon carbide (SiC) ceramic matrix composite (CMC) makes it difficult to accurately model its performance in nuclear applications. The existing models for nuclear grade composite SiC do not account for the microstructural features and how these features can affect the thermal and structural behavior of the cladding and its anisotropic properties. In addition to the microstructural features, the properties of individual constituents of the composites and fiber tow architecture determine the bulk properties. Models for determining the relationship between the individual constituents’ properties and the bulk properties of SiC composites for nuclear applications are absent, although empirical relationships exist in the literature. Here, a hierarchical multiscale modeling approach was presented to address this challenge. This modular approach addressed this difficulty by dividing the various aspects of the composite material into separate models at different length scales, with the evaluated property from the lower-length-scale model serving as an input to the higher-length-scale model. The multiscale model considered the properties of various individual constituents of the composite material (fiber, matrix, and interphase), the porosity in the matrix, the fiber volume fraction, the composite architecture, the tow thickness, etc. By considering inhomogeneous and anisotropic contributions intrinsically, our bottom-up multiscale modeling strategy is naturally physics-informed, bridging constitutive law from micromechanics to meso-mechanics and structural mechanics. The effects that these various physical attributes and thermo-physical properties have on the composite’s bulk thermal properties were easily evaluated and demonstrated through the various analyses presented herein. Since silicon carbide fiber-reinforced SiC CMCs are also promising thermal–structural materials with a broad range of high-end technology applications beyond nuclear applications, we envision that the multiscale modeling method we present here may prove helpful in future efforts to develop and construct reinforced CMCs and other advanced composite nuclear materials, such as MAX phase materials, that can service under harsh environments of ultrahigh temperatures, oxidation, corrosion, and/or irradiation. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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24 pages, 9428 KiB  
Article
An Investigation of Structural Integrity and Dynamic Response of CSB in Case of Slanted Contact with RV Outlet Nozzle
by Pham Quang Hieu and Ihn Namgung
Energies 2024, 17(23), 6070; https://doi.org/10.3390/en17236070 - 2 Dec 2024
Viewed by 332
Abstract
One of the challenges in the current reactor is the gap between the outlet nozzle of the Reactor Vessel (RV) and the outlet opening of the Core Support Barrel (CSB). This gap causes bypass flow, reducing the overall efficiency. To address this issue, [...] Read more.
One of the challenges in the current reactor is the gap between the outlet nozzle of the Reactor Vessel (RV) and the outlet opening of the Core Support Barrel (CSB). This gap causes bypass flow, reducing the overall efficiency. To address this issue, RV and CSB designs were modified with various configurations. Through finite element analysis, this study evaluated the structural integrity and dynamic response of an RV and CSB with these modified designs. The structural integrity was assessed against ASME code to determine the limits of design parameter changes that met code requirements. Additionally, the natural vibration characteristics of the CSB and RV were analyzed to evaluate improvements in the seismic response by modal analysis. The findings revealed that the CSB design in the case of 3-degree slanted contact with the RV outlet nozzle provided the most optimized results. Moreover, modal analysis indicated a substantial enhancement in seismic response, with the dominant CSB mode frequencies increasing by 30%. This shift, from the 15–20 Hz range to the 20–30 Hz range, is especially noteworthy given the concentration of seismic energy within the 1–20 Hz band. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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16 pages, 3099 KiB  
Article
Study of Gas Amplification Impact on Plateau Curve Characteristics in Boron-Coated Proportional Counters
by Yaolong Liu, Zhi Chen, Youjun Huang, Tingfang Luo, Hongliang Zhu and Wenchao Wu
Energies 2024, 17(22), 5740; https://doi.org/10.3390/en17225740 - 16 Nov 2024
Viewed by 546
Abstract
The characteristics of the plateau curve, specifically its length and slope, in boron-coated proportional counters are key performance indicators that impact the detector’s overall performance. Currently, the lack of research on the plateau curve of boron-coated proportional counters is holding back progress in [...] Read more.
The characteristics of the plateau curve, specifically its length and slope, in boron-coated proportional counters are key performance indicators that impact the detector’s overall performance. Currently, the lack of research on the plateau curve of boron-coated proportional counters is holding back progress in engineering design and scientific research. This study harnesses the Diethorn formula, a calculation method for gas gain, to explore the relationship between gas amplification and the plateau curve in a boron-coated proportional counter. Based on the analysis of experimental data, this study proposes methods for improving the characteristics of the plateau curve in a proportional counter, including modifications to the counter structure and gas pressure, along with an evaluation method for assessing the effectiveness of these improvements. First, the causes of the plateau curve in a boron-coated proportional counter are analyzed through a physical process perspective, identifying factors that influence gas amplification and subsequently affect the plateau curve. Building upon this foundation, the Diethorn formula is utilized to explore the effects of structural parameters and gas pressure on gas multiplication. Finally, experimental validation is conducted, resulting in the proposal of three methods for improving the characteristics of the plateau curve and an evaluation method for assessing the effectiveness of improvements. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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20 pages, 3409 KiB  
Article
Development and Verification of a Multi-Physics Transport Code of Molten Salt Reactor Fission Products
by Liang Chen, Liaoyuan He, Shaopeng Xia, Minyu Peng, Guifeng Zhu, Rui Yan, Yang Zou and Hongjie Xu
Energies 2024, 17(21), 5448; https://doi.org/10.3390/en17215448 - 31 Oct 2024
Viewed by 587
Abstract
The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and [...] Read more.
The transport of fission products in molten salt reactors has attracted much attention. However, few codes can completely describe the transport characteristic, though the migration of fission products in the molten salt reactor is essential to estimate the source term, decay heat, and radiation shielding. This study built a program named ThorFPMC (Thorium Fission Products Migration Code) that can handle the multi-physics transport characteristic based on the flow burnup code ThorMODEc (Thorium MOlten Salt Reactor Specific DEpletion Code). A problem-related depletion chain compression method was applied to decrease the order of the solve matrix. The matrix exponential and splitting methods were applied to solve the steady state and transient calculation, respectively. Error analysis showed that for a specific problem, the simplified depletion chain matrix index method could solve the fission products migration equation with an arbitrary time-step with high speed (s) and high precision (10−4); the splitting method could reach a precision of 10−2 level for the full fuel depletion chain, multi-nodes, and transient problems. Compared to the Strang splitting method, the perturbation splitting method has higher precision and less time consumption. In summary, the developed programmer could describe the migration effect of fission products in molten salt reactors, which provides a significant tool for the design of molten salt reactors. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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24 pages, 13719 KiB  
Article
Monte Carlo Modeling of Isotopic Changes of Actinides in Nuclear Fuel of APR1400 Pressurized Water Reactor
by Mikołaj Oettingen and Juyoul Kim
Energies 2024, 17(19), 4864; https://doi.org/10.3390/en17194864 - 27 Sep 2024
Viewed by 527
Abstract
The aim of this paper is to present the isotopic changes in nuclear fuel during the first reactor cycle of the Korean Advanced Power Reactor 1400 (APR1400). The neutron transport and burnup calculations were performed using the Monte Carlo continuous energy burnup code—MCB. [...] Read more.
The aim of this paper is to present the isotopic changes in nuclear fuel during the first reactor cycle of the Korean Advanced Power Reactor 1400 (APR1400). The neutron transport and burnup calculations were performed using the Monte Carlo continuous energy burnup code—MCB. The three-dimensional numerical model consisting of the reactor pressure vessel with core internals was developed using available geometrical and material data as well as the reactor’s operating conditions. The reactor core was divided into 11 axial and 22 radial burnup zones in order to recreate the spatial distribution of the fuel burnup. The isotopic changes due to the nuclear transmutations and decays were calculated in each burnup zone until the desired average burnup of 17.571 GWd/tHMint was reached. The calculations include changes in the boric acid concentration at defined time steps and the burnout of the gadolinia burnable absorber embedded in the nuclear fuel. This study shows the spatial distribution of minor and major actinides at the end of the reactor cycle as well as the depletion of uranium, the build-up of plutonium, and the formation of neptunium, americium, and curium during the reactor’s operation. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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16 pages, 2519 KiB  
Article
Research on Fault Prediction of Nuclear Safety-Class Signal Conditioning Module Based on Improved GRU
by Zhi Chen, Miaoxin Dai, Jie Liu and Wei Jiang
Energies 2024, 17(16), 4063; https://doi.org/10.3390/en17164063 - 16 Aug 2024
Viewed by 613
Abstract
To improve the reliability and maintainability of the nuclear safety-class digital control system (DCS), this paper conducts a study on the fault prediction of critical components in the output circuit of the nuclear safety-class signal conditioning module. To address the issue of insufficient [...] Read more.
To improve the reliability and maintainability of the nuclear safety-class digital control system (DCS), this paper conducts a study on the fault prediction of critical components in the output circuit of the nuclear safety-class signal conditioning module. To address the issue of insufficient feature extraction for the minor offset fault feature and the low accuracy of fault prediction, a predictive model based on stacked denoising autoencoder (SDAE) feature extraction and an improved gated recurrent unit (GRU) is proposed. Therefore, fault simulation modeling is performed for critical components of the signal output circuit to obtain fault datasets of critical components, and the SDAE model is used to extract fault features. The fault prediction model based on GRU is established, and the number of hidden layers, the number of hidden layer nodes, and the learning rate of the GRU model are optimized using the adaptive gray wolf optimization algorithm (AGWO). The prediction performance evaluation metrics include the root mean square error (RMSE), mean absolute error (MAE), mean absolute percentage error (MAPE), and absolute error (EA), which are used for evaluating the prediction results of models such as the AGWO-GRU model, recurrent neural network (RNN) model, and long short-term memory network (LSTM). The results show that the GRU model optimized by AGWO has a better prediction accuracy (errors range within 0.01%) for the faults of the circuit critical components, and, moreover, can accurately and stably predict the fault trend of the circuit. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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12 pages, 8432 KiB  
Article
Assessment of Metal Foil Pump Configurations for EU-DEMO
by Xueli Luo, Yannick Kathage, Tim Teichmann, Stefan Hanke, Thomas Giegerich and Christian Day
Energies 2024, 17(16), 3889; https://doi.org/10.3390/en17163889 - 7 Aug 2024
Viewed by 764
Abstract
It is a challenging but key task to reduce the tritium inventory in EU-DEMO to levels that are acceptable for a nuclear regulator. As solution to this issue, a smart fuel cycle architecture is proposed based on the concept of Direct Internal Recycling [...] Read more.
It is a challenging but key task to reduce the tritium inventory in EU-DEMO to levels that are acceptable for a nuclear regulator. As solution to this issue, a smart fuel cycle architecture is proposed based on the concept of Direct Internal Recycling (DIR), in which the Metal Foil Pump (MFP) will play an important role to separate the unburnt hydrogen isotopes coming from the divertor by exploiting the superpermeation phenomenon. In this study, we will present the assessment of the performance of the lower port of EU-DEMO after the integration of the MFP. For the first time, a thorough comparison of three different MFP (parallel long tubes, sandwich and halo) designs is performed regarding conductance for helium molecules, the pumping speed and the separation factor for deuterium molecules under different physical and geometric parameters. All simulations were carried out in supercomputer Marconi-Fusion with our in-house Test Particle Monte Carlo (TPMC) simulation code ProVac3D because the code had been parallelized with high efficiency. These results are essential for the development of a suitable MFP design in the vacuum-pumping train of EU-DEMO. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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21 pages, 11564 KiB  
Article
Evaluation of Transport–Burnup Coupling Strategy in Double-Heterogeneity Problem
by Yunfei Zhang, Qian Zhang, Yang Zou, Bo Zhou, Rui Yan, Guifeng Zhu, Jian Guo and Ao Zhang
Energies 2024, 17(15), 3792; https://doi.org/10.3390/en17153792 - 1 Aug 2024
Viewed by 666
Abstract
The simulation of fuel composition requires coupled calculations of neutron transport and burnup. It is generally assumed that the neutron flux density and cross-sections remain constant within a burnup step. However, when there are strong absorber poisons present, the reaction rates of the [...] Read more.
The simulation of fuel composition requires coupled calculations of neutron transport and burnup. It is generally assumed that the neutron flux density and cross-sections remain constant within a burnup step. However, when there are strong absorber poisons present, the reaction rates of the absorbers change too rapidly over time, necessitating extremely fine step sizes to ensure computational accuracy, which in turn leads to low computational efficiency. As a type of accident tolerant fuel (ATF), fully ceramic micro-encapsulated (FCM) fuel is a promising new type of nuclear fuel. Accelerated algorithms for burnup calculations of FCM fuel containing gadolinium isotopes have been developed based on the ALPHA code, including the projected predictor–corrector (PPC), the log-linear rate (LLR), and the high-order predictor–corrector (HOPC) methods (including CE/LI, CE/QI, LE/LI, and LE/QI). The performances of different algorithms under the two forms of Gd2O3 existence were analyzed. The numerical results show that the LE/QI method performs the best overall. For Gd2O3 existing in both forms, the LE/QI algorithm can maintain accuracy with a burnup step size of up to 1.0 GWd/tU, keeping the infinite multiplication factor kinf within 100 pcm, and it exhibits high accuracy in simulating the atomic number densities of Gd-155 and Gd-157 throughout the burnup process. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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14 pages, 5194 KiB  
Article
Development of a MELCOR Model for LVR-15 Severe Accidents Assessment
by Alain Flores y Flores, Guido Mazzini and Antonio Dambrosio
Energies 2024, 17(14), 3384; https://doi.org/10.3390/en17143384 - 10 Jul 2024
Viewed by 656
Abstract
LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by [...] Read more.
LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for modelling the progression of a severe accident (SA) in light-water nuclear power plants and estimating the behaviour of the reactor under SA conditions. The model was built by collecting information about the LVR-15. Since the research reactor can have different core configurations according to the location of the core components, the core configuration with the most fuel (hottest campaign K221) was selected. Then, to create the radial nodalisation, the details of the core components were obtained and grouped in five radial rings and 27 axial levels. The simulation was run with the boundary conditions collected from campaign K221, and the results were compared with the reference values of the campaign with a negligible percentage of error. For the coolant inlet and outlet temperature, the reference values were 318.18 K and 323.5 K, respectively, while for the simulation, the steady state reached 319 K for the inlet temperature and 324 K for the outlet temperature. Additionally, the cladding temperature of the hottest assembly was compared with the reference value (353.72 K) and the steady-state simulation results (362 K). In future work, different transients leading to severe accidents will be simulated. When simulating the LVR-15 reactor with MELCOR, specific attention is required for the aluminium-cladded fuel assemblies, as the model requires some assumptions to cope with the phenomenological limitations. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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12 pages, 2366 KiB  
Article
RFP-MSR Hybrid Reactor Model for Tritium Breeding and Actinides Transmutation †
by Stefano Murgo, Chiara Bustreo, Marco Ciotti, Guglielmo Lomonaco, Francesco Paolo Orsitto, Roberto Piovan, Nicola Pompeo, Giovanni Ricco, Marco Ripani and Fabio Panza
Energies 2024, 17(12), 2934; https://doi.org/10.3390/en17122934 - 14 Jun 2024
Cited by 1 | Viewed by 994
Abstract
The studies on the development of fusion–fission hybrid reactors (FFHR) have gained consensus in recent years as an intermediate step before fusion energy. This work proposes a possible approach to FFHRs based on the coupling of a Reversed Field Pinch fusion machine and [...] Read more.
The studies on the development of fusion–fission hybrid reactors (FFHR) have gained consensus in recent years as an intermediate step before fusion energy. This work proposes a possible approach to FFHRs based on the coupling of a Reversed Field Pinch fusion machine and a Molten Salt Subcritical fission test bed. The proposed test bed is characterized by the coexistence of a fast-neutron fission core and a dedicated thermal-neutron zone, allowing the performing of tritium breeding and actinides transmutation studies. The neutronic design solutions and the results obtained by the irradiation of FLiBe salt (inside the thermal-neutron zone) and of an actinide target (inside the core) are shown. The outcomes of the analysis reveal the potential of FFHR systems as breeding/burner systems. In particular, the results regarding tritium breeding are very encouraging as the system is demonstrated to be able to reach a very high Tritium Breeding Ratio. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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Review

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15 pages, 3053 KiB  
Review
Matching Analysis of Technical Parameters and Safety Standards for Nuclear Replacement of Coal-Fired Units
by Dongwang Zhang, Tuo Zhou, Zhihong Liu, Hairui Yang, Rushan Bie and Man Zhang
Energies 2024, 17(22), 5583; https://doi.org/10.3390/en17225583 - 8 Nov 2024
Viewed by 492
Abstract
In the context of the growing share of renewable energy and the impending decommission of a large number of coal-fired units, nuclear energy is the only green energy that can replace coal power for a stable, clean, efficient, and large-scale power supply. This [...] Read more.
In the context of the growing share of renewable energy and the impending decommission of a large number of coal-fired units, nuclear energy is the only green energy that can replace coal power for a stable, clean, efficient, and large-scale power supply. This article compares the differences between coal power and nuclear power in terms of thermal system, thermal cycle, turbine parameters, and safety. It discusses the possibility of replacing the boiler of a coal power plant with nuclear power, that is, replacing the boiler of a coal power plant with a nuclear reactor for generation/heating/cogeneration. For coal-fired units with similar capacity that do not use a reheat cycle (at or below high pressure) and nuclear power units (such as a high-temperature gas-cooled reactor), as well as coal-fired units with a reheat cycle (ultra-high pressure and above) and nuclear power units (such as pressurized water reactors), there are great differences in steam parameters. In terms of steam turbines, the size of nuclear power units is relatively larger, requiring additional dehumidification measures. In addition, the safety factors and management methods considered in the site selection, construction, and operation of nuclear power plants are more stringent and complex, and comprehensive analysis and evaluation are needed in aspects such as waste treatment and accident emergency response. Except for the relevant provisions of the American Society of Mechanical Engineers code for pressure vessels, nuclear power units are not compatible with coal-fired units in terms of safety standards. Therefore, considering comprehensively, it is believed that the scheme of nuclear power replacing coal-fired units for power generation/heating/cogeneration program is not feasible at present. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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15 pages, 692 KiB  
Review
Review on the Application of Living PSA in Nuclear Power
by Min Zhang, Zhao Xu, Guichang Zhang, Binbin Wang, Bin Zhang and Yilong Liu
Energies 2024, 17(22), 5578; https://doi.org/10.3390/en17225578 - 8 Nov 2024
Viewed by 502
Abstract
With the increasing standards of safety management in nuclear power plants, Living Probabilistic Safety Assessment (Living PSA) technology has begin to play an increasingly important role in their operation. This paper aims to provide an overview of the application and development of Living [...] Read more.
With the increasing standards of safety management in nuclear power plants, Living Probabilistic Safety Assessment (Living PSA) technology has begin to play an increasingly important role in their operation. This paper aims to provide an overview of the application and development of Living Probabilistic Safety Assessment (Living PSA) technology in nuclear power plant safety monitoring and risk assessment, examining the key technologies and future challenges. Initially, we summarize the current safety needs in regard to nuclear power, examine the policy on configuration risk management technology for nuclear power plants, and outline its importance and development process in nuclear power plant safety management. Subsequently, we discuss the basic principle of Living PSAs and the working method of risk monitoring based on Living PSAs, including information monitoring data collection, online identification, real-time model updating, and risk calculation. Within the Living PSA framework, model development is not merely about creating a theoretical or static representation; it is a dynamic and ongoing process that involves a deep understanding and precise simulation of the behavior of nuclear power plant systems and components. This represents the main research efforts in Living PSAs at present. Additionally, this paper identifies the key technologies of Living PSAs in an in-depth manner, such as the reliability-model-updating technology and model building in dynamic reliability analyses, including the fault tree model, multi-layer flow model, GO-FLOW model etc. The paper lists the work of some scholars in this area in recent years, which helps readers and researchers to clearly understand the current progress of Living PSA technologies in terms of model establishment and updating. Finally, the paper summarizes the challenges and future development of Living PSA and emphasizes the possible problems in data quality, human factor engineering, and the development of Living PSA technologies in the future. In the future, Living PSAs will provide more solid support for the realization of safer and more economical methods of operating nuclear power plants. Full article
(This article belongs to the Special Issue Advanced Technologies in Nuclear Engineering)
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Planned Papers

The below list represents only planned manuscripts. Some of these manuscripts have not been received by the Editorial Office yet. Papers submitted to MDPI journals are subject to peer-review.

Title: Development of AI-based image reconstruction model for Partial Defects Verification in Nuclear Fuel Assemblies
Authors: Jae Joon Ahn
Affiliation: Division of Data Science, Yonsei University, Wonju 26493, Republic of Korea
Abstract: The increase in nuclear power plants for carbon neutrality has emphasized the importance of managing high-radiation spent nuclear fuel (SNF). Traditional inspection methods, such as Gamma Emission Tomography (GET), have limitations in detecting partial defects within SNF assemblies. This study aims to enhance defect detection accuracy by optimizing AI-based image classification algorithms. Using emission tomography image data from 3x3 nuclear fuel assemblies, we compared the performance of neural network models (AlexNet, ResNet, SENet) and tree-based ensemble models (XGBoost, Random Forest, LightGBM). Our results show that neural network models, particularly ResNet and SENet, achieve superior classification accuracy with limited training data. SENet, in particular, demonstrates high performance with fewer samples, indicating its effectiveness in defect detection with minimal data. Tree-based models like XGBoost and LightGBM also exhibit high accuracy but are slightly lower than neural networks. In conclusion, AI-based classification systems, especially those utilizing advanced neural networks, can significantly improve the inspection and management of SNF, ensuring safety and compliance in nuclear energy operations. Future work should explore these methodologies on larger configurations beyond the 3x3 assemblies to further validate their effectiveness.

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