Progress in Nuclear Energy最新文献

{"title":"Modelling and assessment of thermophysical properties of Am-bearing fuels for transmutation purposes in fast reactors","authors":"M. Di Gennaro ,&nbsp;A. Magni ,&nbsp;M. Guarnieri ,&nbsp;D. Pizzocri ,&nbsp;L. Luzzi ,&nbsp;C. Guéneau ,&nbsp;P. Van Uffelen","doi":"10.1016/j.pnucene.2025.105698","DOIUrl":"10.1016/j.pnucene.2025.105698","url":null,"abstract":"<div><div>Minor Actinides (MAs) are produced inside nuclear reactor fuels under irradiation and are responsible of large part of the waste radiotoxicity. Partitioning and Transmutation (P&amp;T) is identified as the strategy that can relax constraints on geological disposal, since transmutation of MAs via critical or subcritical fast spectrum irradiation facilities like MYRRHA is a valid path to reduce the waste hazard. One of the focuses of the PATRICIA Project is the development of fuel performance codes towards Am-bearing fuels and the computational study of their behaviour under irradiation, supporting the licencing and development of advanced designs of the MYRRHA reactor. This work focuses on the modelling of thermophysical properties of transmutation-type fuel concepts for fast reactor application, specifically of melting temperature, specific heat capacity and thermal conductivity of (U,Pu,Am)O<span><math><msub><mrow></mrow><mrow><mn>2</mn><mo>−</mo><mi>x</mi></mrow></msub></math></span> and (U,Am)O<span><math><msub><mrow></mrow><mrow><mn>2</mn><mo>−</mo><mi>x</mi></mrow></msub></math></span>. The modelling activity involves the latest atomic-scale/CALPHAD data on Am-bearing fuels achieved in the framework of PATRICIA. Developed models are verified and validated against separate-effect data, and implemented in the TRANSURANUS fuel performance code to simulate a past irradiation experiment (HEDL P-19 power-to-melt transient test) for the evaluation of the impact of new models on the pin performance and their integral validation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105698"},"PeriodicalIF":3.3,"publicationDate":"2025-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143654532","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Optimizing the local-scale turbulence-induced dispersion of LAPMOD at a Chinese nuclear power plant with heterogeneous topography and irregular building layouts","authors":"Li Yang ,&nbsp;Sheng Fang ,&nbsp;Zhaoyang Wang ,&nbsp;Cunyou Wang ,&nbsp;Xinwen Dong ,&nbsp;Xinpeng Li ,&nbsp;Yixue Chen","doi":"10.1016/j.pnucene.2025.105720","DOIUrl":"10.1016/j.pnucene.2025.105720","url":null,"abstract":"<div><div>This study proposes a composite correction scheme to address the inadequate turbulence-induced dispersion of the Lagrangian Particle Model (LAPMOD) at a nuclear power plant site with complex terrains and irregular buildings. The proposed scheme combines both background and local optimization of the turbulence intensity, which are overall lower bound settings and local correction in building zones. The feasibility of the correction is comprehensively assessed by comparing multiple LAPMOD's simulations with measurements from two wind tunnel experiments of a Chinese nuclear power plant. Furthermore, a comprehensive sensitivity analysis is conducted to optimize the mixing length parameterization in the surface layer of the proposed scheme. The results show that the correction scheme significantly enhances the agreement between simulations and measurements, with respect to both spatial plume coverage and quantitative metrics. It effectively mitigates long-range overestimations in the plume center, reduces deviations at axial sites, and accurately reproduces the vertical concentration profile at the main mountain and building-area boundary. The quantitative metrics confirm the excellent performance of the correction scheme, achieving the acceptance criteria of all four quantitative statistical metrics. The sensitivity analysis reveals that the optimal mixing length is the ensemble mean of two classic schemes, enhancing applicability of the proposed scheme in practice.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105720"},"PeriodicalIF":3.3,"publicationDate":"2025-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143654531","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Study on modular design methodology of marine SMR system based on fuzzy hierarchical clustering and improved genetic algorithm","authors":"Zhihong Tang ,&nbsp;Liyuan Wang ,&nbsp;Shusheng Guo ,&nbsp;Guopeng Liang ,&nbsp;Wenjun Zhang ,&nbsp;Lide Zhang ,&nbsp;Ming Rui ,&nbsp;Guan Guan ,&nbsp;Yunlong Wang","doi":"10.1016/j.pnucene.2025.105739","DOIUrl":"10.1016/j.pnucene.2025.105739","url":null,"abstract":"<div><div>Marine Small Modular Reactors (MSMR) integrate SMR technology with ship technology, offering unique value in meeting the energy demands of the open ocean and remote islands. However, the design and construction of MSMR face challenges such as space constraints, complex system integration, and the need to adapt to advanced ship modular construction technologies. Therefore, efficient modular partitioning methods are required to enhance overall efficiency and reliability. The module partitioning of MSMR systems takes into account multiple factors and is a combinatorial optimization problem with performance constraints. This study aims to reflect the internal structure of the system hierarchical tree, provide clear guidance for module partitioning, and improve the computational efficiency of solving combinatorial problems. This paper propose a module division and optimization method for MSMR systems based on fuzzy hierarchical clustering and a genetic algorithm. Initially, the components of the small modular reactor power plant system are clustered into modules of different levels using fuzzy hierarchical clustering. Subsequently, a genetic algorithm is employed to solve the combinatorial optimization problem of the module division scheme, resulting in the optimal division scheme. The feasibility and effectiveness of the method are verified through the modular case of the Radioactive Waste Gas System (WGS). This method can provide guidance for the modularization design of the entire ocean modular reactor system. The method provided in this article can provide a research foundation for future modular design of MSMR and improve design efficiency.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105739"},"PeriodicalIF":3.3,"publicationDate":"2025-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143654570","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical study on the thermal hydraulic effect of flow blockage in liquid metal cooled reactor core rod bundles","authors":"Lingping Song ,&nbsp;Jiahao Zhao ,&nbsp;Shiqi Wang ,&nbsp;Bin Han ,&nbsp;Xiaoliang Zhu ,&nbsp;Bao-Wen Yang ,&nbsp;Aiguo Liu ,&nbsp;Shenghui Liu ,&nbsp;Junlin Huang","doi":"10.1016/j.pnucene.2025.105743","DOIUrl":"10.1016/j.pnucene.2025.105743","url":null,"abstract":"<div><div>Liquid metal reactor has become one of the most promising types of Gen IV nuclear energy systems due to its excellent safety and thermal hydraulic characteristics. Flow blockage is a severe thermal accident especially in liquid metal reactors compared to other reactors. When fuel assembly is blocked, decrease of coolant will cause the temperature to dramatically rise in the cladding and unexpected flow downstream the blocks. The safety of the reactor will be affected. In this paper, a lead-cooled 7 pin bare rod bundle model was established based on CFD to simulate the flow behaviors under blockage accidents. Cases with different geometries including block shape, axial length, block area and boundary conditions were set to investigate the effect of flow blockage. The temperature, pressure and velocity field downstream the blocks were obtained. It was found that the boundary conditions and geometry parameters have coupled effect on the flow performance of the blockage. A critical axial length of blocks was found to make the effect of blockage reach the peak. The research provides a fundamental basis for the design and safe operation of liquid metal reactor.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105743"},"PeriodicalIF":3.3,"publicationDate":"2025-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143642528","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Physics-Informed Neural Networks for the safety analysis of nuclear reactors","authors":"Piero Baraldi ,&nbsp;Camilla Battistini ,&nbsp;Federico Antonello ,&nbsp;Jacopo Buongiorno ,&nbsp;Enrico Zio","doi":"10.1016/j.pnucene.2025.105745","DOIUrl":"10.1016/j.pnucene.2025.105745","url":null,"abstract":"<div><div>This work explores the development of surrogate models for estimating the evolution of quantities of interest during nuclear reactor accident scenarios. Physics-Informed Neural Networks (PINNs) offer a promising surrogate modelling approach because they allow integrating laws of physics and domain knowledge into traditional Neural Network (NN) surrogates. Specifically, the proposed solution incorporates an additional term in the PINN loss function to enforce physics-based constraints in correspondence of allocation points, which are randomly sampled points whose corresponding target output is not known. As a result, accuracy of the estimation of the quantities of interest and their adherence to the laws of physics are improved. Applications to a synthetic case study and to the response of a nuclear microreactor system during a Loss of Heat Sink scenario confirm that the developed surrogate model based on PINN with allocation points improves the estimation accuracy with respect to other state-of-the-art methods.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105745"},"PeriodicalIF":3.3,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143642527","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Optioneering in nuclear ion exchange resin disposal","authors":"C.M. Atkinson ,&nbsp;T.J. Robshaw ,&nbsp;G.D. Walker ,&nbsp;M.J.D. Rushton ,&nbsp;S.C. Middleburgh ,&nbsp;W.E. Lee ,&nbsp;M.D. Ogden","doi":"10.1016/j.pnucene.2025.105719","DOIUrl":"10.1016/j.pnucene.2025.105719","url":null,"abstract":"<div><div>Spent ion exchange resins from nuclear facilities make up 4 % of all solid waste from that industry, and with no current disposal pathway, represent an orphan waste stream. Currently spent ion exchange resins are stored in vaults where they slowly degrade, producing dangerous secondary waste that is harder to dispose of.</div><div>In this work the most viable disposal options have been assessed using Multi Criteria Decision Analysis (MCDA). Several methodologies were appraised, and a modified version of the Analytical Hierarchy Process (AHP) was chosen. Disposal options identified were shortlisted by the application of threshold criteria. Using AHP, eight Key Performance Indicators (KPIs) were defined and grouped into three overarching criteria: Engineering, Economics and Environmental. The KPIs and criteria were subjected to a ‘pairwise rating process’ to assess their relative importance, which was then used to calculate weightings. These weightings were used to prioritise the raw scores each disposal route had received for the KPIs giving rankings for each disposal option. These were analysed using an uncertainty assessment, employing additional indicators to give an uncertainty percentage. A sensitivity analysis was conducted which changed the weightings to assess the impact this could have on the final rankings.</div><div>Vitrification was found to be the most viable option, achieving good scores in all categories. Cementation and Plastic Solidification were also found to be promising, offering a more cost-effective option. A control disposal option, designed to emulate the current UK strategy of storage pending treatment, was included in the analysis. As expected, this disposal option scored poorly, ranking 11th out of 14 demonstrating that the current approach is unsustainable. Several viable alternatives are suggested with further laboratory and economic studies that would allow for deployment of the chosen disposal options.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105719"},"PeriodicalIF":3.3,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143636879","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Non-destructive evaluation and machine learning methods for inspection of spent nuclear fuel canisters: A state-of-the-art review","authors":"Bozhou Zhuang ,&nbsp;Anna Arcaro ,&nbsp;Bora Gencturk ,&nbsp;Ryan Meyer ,&nbsp;Assad Oberai ,&nbsp;Anton Sinkov ,&nbsp;Morris Good","doi":"10.1016/j.pnucene.2025.105697","DOIUrl":"10.1016/j.pnucene.2025.105697","url":null,"abstract":"<div><div>Nuclear energy is among the cleanest and most efficient energy sources currently available. The operation of nuclear power plants (NPPs) produces large amounts of high-level radioactive waste known as spent nuclear fuel (SNF). Currently, large amounts of SNF is stored in dry cask storage systems (DCSSs) for extended interim storage until a permanent disposal solution becomes available. During the extended interim storage, the DCSS, particularly the SNF canisters, may degrade and abnormal conditions may occur. Therefore, non-destructive evaluation (NDE) and machine learning (ML) approaches are necessary for inspection of SNF canisters. This paper presents a state-of-the-art review of literature by summarizing recent progress made on the applications of NDE and ML for inspection of SNF canisters. Sixteen NDE methods are examined and compared: visual inspection, ultrasonic guided waves (UGWs), laser-based approaches, acoustic emission (AE), eddy current testing (ECT), non-invasive acoustic sensing, dynamic modal testing, cosmic ray muons tomography, neutron imaging, gamma rays detection, fiber optical sensors, through-wall communications, X-ray computed tomography (CT), vibrothermography, monoenergetic photon sources, and surface acoustic wave (SAW) sensors. The technology readiness level (TRL) for each method is assessed and compared. Recent publications on ML-enhanced visual inspection, AE, non-invasive acoustic sensing, dynamic modal testing, and neutron imaging for SNF canisters are summarized and future research needs are identified. This review article provides a convenient reference on the state-of-the-art applications of NDE and ML methods for inspection of SNF canisters.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105697"},"PeriodicalIF":3.3,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143636877","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Study on thermal-hydraulic performance of airfoil fin PCHE using vortex generators with different arrangement parameters for SCO2 Brayton cycle","authors":"Yangguang Zhang ,&nbsp;Zhen Li ,&nbsp;Daogang Lu ,&nbsp;Qiong Cao ,&nbsp;Henghua Liang","doi":"10.1016/j.pnucene.2025.105724","DOIUrl":"10.1016/j.pnucene.2025.105724","url":null,"abstract":"<div><div>The performance of the airfoil fin printed circuit heat exchanger (PCHE) is of great importance for the efficiency and size scale of Generation IV nuclear energy systems coupled to supercritical carbon dioxide (SCO₂) Brayton cycles. Arranging vortex generator (VG) inside the airfoil fin channel could effectively enhance the comprehensive performance of PCHE, however, related studies are still limited. In particular, the study of the thermal-hydraulic characteristics of SCO₂ in the PCHE channel with different arrangement parameters of the VG, which is important for the optimization of the PCHE performance, is very lacking. Therefore, it is of great significance to study the influence of parameters such as vertical spacing (L_a,VG) and horizontal spacing (L_b,VG) between the VG and the fin, the angle of attack (θ) and offset direction of the VG on the flow and heat transfer performance of the SCO₂ in the PCHE flow channel. This study compares the flow and heat transfer performance of the channel with different arrangement parameters of VG, and reveals the flow and heat transfer mechanism of SCO₂ in the channel. The results show that with the increase of L_a,VG, the thermal performance of the PCHE channel is weakened and then strengthened, and the flow performance is gradually weakened. With the increase of L_b,VG, the thermal performance of the PCHE channel is enhanced and then weakened, the flow performance is weakened, then enhanced and finally weakened. With the increase of θ, the thermal performance of the PCHE channel is gradually enhanced, while the flow performance is gradually weakened. The thermal performance of the PCHE channel is best when the VG deviation direction is different on both sides of the fin. Field synergy principle, entropy production principle, distribution of turbulent kinetic energy, distribution of vortex structure, and properties of SCO₂ can well explain these. In addition, among the different VG arrangement parameters, L_a,VG of 0.7 mm, L_b,VG of 3 mm, θ of 45°, and the different direction of VG deviation on both sides of the airfoil fins are recommended.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105724"},"PeriodicalIF":3.3,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143636878","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Heat transfer characteristics and design optimization of compact heat exchanger with liquid lead-bismuth eutectic and sCO2 as working fluids","authors":"Zhiyuan Luo,&nbsp;Gen Li","doi":"10.1016/j.pnucene.2025.105740","DOIUrl":"10.1016/j.pnucene.2025.105740","url":null,"abstract":"<div><div>The Printed Circuit Heat Exchanger (PCHE) is one of the potential intermediate heat exchangers (IHXs) for nuclear reactors characterised by its high efficiency, high power density, and high endurance in extreme conditions. The use of PCHE as an IHX of a liquid lead-bismuth eutectic (LBE) reactor coupled with the supercritical carbon dioxide (sCO₂) Brayton cycle system can enhance its compactness and safety. However, the research on the thermal-hydraulic characteristics of PCHE with LBE and sCO₂ as working fluids is scarce. In this research, the flow and heat transfer of LBE and sCO₂ in straight, zigzag, and airfoil fins channels were studied by numerical method. The results indicated the zigzag channel had a larger Nusselt number (<em>Nu</em>) and friction factor (<em>f</em>) with significant fluctuations along the channel due to the periodic bending structures. While the stable <em>Nu</em> and <em>f</em> were observed in the straight and airfoil fins channels. As the inlet velocity rose, the heat transfer increased while the friction loss decreased. The larger inlet temperature of sCO₂ resulted in lower <em>Nu</em> and higher <em>f</em>, while the inlet temperature had an unobvious impact on LBE. New correlations of sCO₂ and LBE were developed across an extensive range of Reynolds number. Moreover, the economic viability of three channel PCHEs was compared by multi-objective optimization method and cost analysis, revealing that the annual total cost of the airfoil fins channel PCHE was 9.7% and 4.7% lower than that of the straight and zigzag, respectively.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105740"},"PeriodicalIF":3.3,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143619877","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Progress in understanding the formation and deposition behaviors of corrosion and oxidation products in lead-cooled fast reactors","authors":"Zhengang Duan ,&nbsp;Hongchen Xiang ,&nbsp;Pengrui Qiao ,&nbsp;Qinglong Wen ,&nbsp;Huiping Ouyang","doi":"10.1016/j.pnucene.2025.105742","DOIUrl":"10.1016/j.pnucene.2025.105742","url":null,"abstract":"<div><div>The Lead-cooled fast reactor (LFR) is regarded as a promising fourth-generation advanced power reactor. However, the development of LFR is limited by the corrosion problems of structural materials. Corrosion and oxidation products can deposit on the heat transfer surface during the migration process of the coolant, leading to heat transfer degradation and posing a serious threat to reactor operation safety. In recent years, researchers have conducted extensive research on corrosion protection, as well as the migration, deposition, and impact of corrosion and oxidation products. This paper focuses on oxygen control technology, summarizes the current corrosion protection technologies for lead-cooled fast reactors, introduces the formation mechanisms of common corrosion and oxidation products in lead-cooled fast reactors, briefly summarizes the migration and deposition mechanisms of corrosion and oxidation products, and finally puts forward suggestions for further research based on the current research progress.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105742"},"PeriodicalIF":3.3,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143619889","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}