Progress in Nuclear Energy最新文献

{"title":"Investigation on cross flow induced tube bundle vibration and tube-support contact state","authors":"Zhenqin Xiong,&nbsp;Shuo Wang,&nbsp;Ruiqi Kang,&nbsp;Jing Liu,&nbsp;Linpeng Shi,&nbsp;Shenjie Gong","doi":"10.1016/j.pnucene.2025.106068","DOIUrl":"10.1016/j.pnucene.2025.106068","url":null,"abstract":"<div><div>Wear due to non-uniform cross flow induced vibration (FIV) challenging the integrity of tubes and fuel rods in nuclear reactors depends on the vibration trajectory and the contact state between the tube and the tube support plate. To reveal the characteristics of vibration and contact state of tubes suffering non-uniform cross flow, experimental investigation on the vibration response of a three-span 7x11 tube bundle subjected to cross-flow from a one-third opening baffle and a baffle with uniform holes has been carried out. Visualization measurement techniques have been developed to obtain the vibration trajectory of the tube in the middle of the first row and the micro-meter sale gap between the tube and the support plate at the top end. Numerical simulations were also conducted to analyze the flow field characteristics within the tube bundle formed by the one-third opening inlet flow. The results indicate that the tube bundle subjected to flow from a one-third open inlet is more prone to fluid-elastic instability compared to that with a uniformly perforated inlet. However, in the high velocity ranging from 1.52 m/s to 2.77 m/s before FEI, the case with uniform-hole baffle has higher vibration transverse displacement at the top end than the corresponding case with one-third opening baffle. The gap between the tube C1 (middle tube in the first row) and 2#TSP, and the top end trajectory of tube C1 is systematically analyzed. The contact states between this tube and the 2#TSP indicate that the tube stably leaning on the TSP is prone to adhesive wear before FEI for non-uniform flow case, while the tube experiencing serious displacement at contact location is prone to sliding wear before the FEI for uniform flow.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106068"},"PeriodicalIF":3.2,"publicationDate":"2025-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145220476","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 of Droplet motion behavior in corrugated-plate dryers","authors":"Aoxiang Qin,&nbsp;Haoyang Liao,&nbsp;Wenjie Ma,&nbsp;Sichao Tan,&nbsp;Jiming Wen,&nbsp;Ruifeng Tian","doi":"10.1016/j.pnucene.2025.106048","DOIUrl":"10.1016/j.pnucene.2025.106048","url":null,"abstract":"<div><div>As a critical dehumidification device in the steam systems of the nuclear industry, the separation efficiency of corrugated-plate dryers influences steam quality and operational safety directly. The theoretical model enables rapid adjustment of parameters in the design optimization of CPDs, offering significantly higher efficiency compared to experimental measurements and CFD simulations. This study establishes a droplet motion model based on the force analysis of the droplet at different sections of CPDs. The model integrates the Rosin-Rammler droplet size distribution with a wall collision detection algorithm, and develops a predictive program for the separation efficiency of CPDs. The predicted results of the program show good consistency with experimental data, with a maximum deviation of 5.4 % in separation efficiency. The influence mechanism of inlet parameters and structural parameters on the separation efficiency of the CPD was studied based on the movement trajectory of droplets in the channel.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106048"},"PeriodicalIF":3.2,"publicationDate":"2025-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145158990","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":"Duty cycle reactivity control and xenon spatial oscillation stability analysis of a boron-free small modular reactor core","authors":"Bright Madinka Mweetwa ,&nbsp;Marat Margulis","doi":"10.1016/j.pnucene.2025.106062","DOIUrl":"10.1016/j.pnucene.2025.106062","url":null,"abstract":"&lt;div&gt;&lt;div&gt;One of the challenges associated with operating a boron-free Small Modular Reactor (BFSMR) is increased use of rod cluster control assemblies (RCCAs) for duty cycle operations and power manoeuver, and burnable poisons for excess reactivity control. Frequent RCCA movement may result in increased xenon spatial oscillations. In this work four LPs with similar thermal output of 1429.51 MW were assessed to determine a candidate LP for the BFSMR core. CASMO4, ROSA and SIMULATE3 were used for fuel assembly modelling, core loading and 3D core simulation and assessment. The LP with 32 feed fuel assemblies and a maximum of 56 &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;G&lt;/mi&gt;&lt;mi&gt;d&lt;/mi&gt;&lt;/mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/msub&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;/mrow&gt;&lt;mn&gt;3&lt;/mn&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; doped fuel pins in a fuel assembly was adopted as the best LP. A RCCA insertion pattern for duty cycle operation of the candidate LP and the power-dependent rod insertion limit (PDIL) for power manoeuver have been established. In addition, free-xenon stability of core and RCCA movement-initiated xenon transient were assessed. The peak linear heat rate (&lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;), enthalpy rise hot channel factor (&lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mrow&gt;&lt;mo&gt;Δ&lt;/mo&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and the heat flux hot channel factor (&lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mi&gt;q&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; were used as a criterion for acceptance of RCCA insertion pattern at each burnup step and each power level associated with power manoeuver. The target &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mrow&gt;&lt;mo&gt;Δ&lt;/mo&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mi&gt;q&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; were set at 42.12 kW/m, 1.65, and 2.6, respectively. The maximum values for &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mrow&gt;&lt;mo&gt;Δ&lt;/mo&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mi&gt;q&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; associated with the RCCA insertions insertion pattern were found to be 38.2 kW/m, 1.608, and 2.38. The maximum &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mrow&gt;&lt;mo&gt;Δ&lt;/mo&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mtext&gt;,&lt;/mtext&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mi&gt;q&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; values associated with power manoeuver for the power range 80%–100% were found to be 33.01 kW/m, 1.62, and 2.26 respectively. These values were associated with the 8.5 MWd/kg burnup step – a burnup step at which &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mi&gt;G&lt;/mi&gt;&lt;mi&gt;d&lt;/mi&gt;&lt;/mrow&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/msub&gt;&lt;msub&gt;&lt;mi&gt;O&lt;/mi","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106062"},"PeriodicalIF":3.2,"publicationDate":"2025-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145158992","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":"The decay heat distribution calculation in molten salt reactor experiment based on the nuclide flow-transfer-burnup coupling method","authors":"Zhenghao Xu ,&nbsp;Guifeng Zhu ,&nbsp;Liang Chen ,&nbsp;Shuyang Jia ,&nbsp;Changqing Yu ,&nbsp;Yunfei Zhang ,&nbsp;Yang Zou ,&nbsp;Hongjie Xu","doi":"10.1016/j.pnucene.2025.106060","DOIUrl":"10.1016/j.pnucene.2025.106060","url":null,"abstract":"<div><div>The liquid-fueled molten salt reactor (MSR), unique among Generation IV systems, presents novel challenges due to its circulating fuel's multi-phase decay heat distribution. This study develops a nuclide flow-transfer-burnup coupling method to analyze spatiotemporal decay heat characteristics in MSRs. By analyze 8 MW Molten Salt Reactor Experiment (MSRE) primary loop, we find that, Short-lived nuclides (1s-1min half-life) contribute ∼40 % of equilibrium decay heat but rapidly decay when leaving the core, leading to flow-dependent spatial non-uniformity; While most decay heat (∼70 %) remains in salt, long-lived insoluble nuclides accumulate at high-surface-area components (e.g., heat exchangers); After reactor shutdown, salt-phase decay heat drops rapidly while wall-deposited nuclides maintain long-term heat output. Our framework enables accurate decay heat analysis under all operational conditions, revealing significant spatial heterogeneity both within and beyond the fuel salt (wall phase, gas phase, gas-remove systems). These findings are crucial for decay heat removal system design. Current uncertainties in bubble dynamics and mass transfer coefficients require further experimental validation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106060"},"PeriodicalIF":3.2,"publicationDate":"2025-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145158993","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":"Establishing a relation between the energy parameters of the research impulse graphite reactor and the tested fuel assembly with a block structure","authors":"Galina Vityuk ,&nbsp;Vladimir Vityuk ,&nbsp;Ruziya Kelsingazina ,&nbsp;Nurzhan Mukhamedov ,&nbsp;Ruslan Irkimbekov","doi":"10.1016/j.pnucene.2025.106052","DOIUrl":"10.1016/j.pnucene.2025.106052","url":null,"abstract":"<div><div>Establishing a connection between the parameters of the fuel assembly (FA) tested at the research reactor and the parameters of the reactor itself is an important step in the procedure for preparing demonstration experiments, the purpose of which is to obtain a set of data on the processes occurring in the core of a nuclear reactor in transient and emergency modes. To set up the research reactor operating mode in a demonstration experiment, it is necessary to know reliably the ratio between the power in the reactor and in the tested FA. In this article, as part of the development of the methodology for the preparation and conduct of reactor tests at Impulse Graphite Reactor (IGR), three independent methods for determining the energy parameters of reactor test facilities with the FA having a block structure have been considered and tested in a series of three calibration reactor experiments and a set of computational studies. The ratio of specific power density in the FA and at the IGR has been determined by the thermal balance method based on the results of measuring the temperature of the structural elements, the gamma spectrometric method based on measurements of the content of fission products in the fuel composition after irradiation, and theoretical computational studies based on the use of computer modeling tools. The resulting ratio value is in the range of (7.73 … 7.78)⋅10⁻⁷ (kJ/g(UO₂))/kJ_IGR. The results of the application of three independent methods are in good agreement with each other, which indicates the absence of contradictions between them and the possibility of using them in the preparation and implementation of experimental testing programs for the reactor fuel at the research.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106052"},"PeriodicalIF":3.2,"publicationDate":"2025-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145118866","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":"An enhanced factor model for coupled conductive - radiative heat transfer at the particle scale in pebble beds","authors":"Hao Wu ,&nbsp;Shuang Hao ,&nbsp;Fang Liu ,&nbsp;Yang Liu ,&nbsp;Fenglei Niu ,&nbsp;Jiyuan Tu","doi":"10.1016/j.pnucene.2025.106061","DOIUrl":"10.1016/j.pnucene.2025.106061","url":null,"abstract":"<div><div>In nuclear pebble beds, heat transfer is a complex physical process involving three primary mechanisms: particle-particle conduction, conduction through the fluid film between contact pairs (particle-fluid-particle conduction), and particle-particle thermal radiation. The enhanced factor model, a novel approach for particle-scale simulations, was developed to efficiently address these combined phenomena. The model incorporates the contribution of thermal radiation directly into the particle conduction equation for contacting particles. This is achieved through an enhanced factor that is approximated using a Taylor polynomial. A key advantage of this method is its ability to bypass complex and computationally expensive procedures such as thermal ray tracing for view factor calculations. This efficiency, combined with the use of a sub-cell radiation model (SCM), allows for the performance of particle-scale simulations of conduction and radiation heat transfer in large-scale pebble beds. The enhanced factor model has been validated against experimental data, showing general agreement and demonstrating its effectiveness as a tool for analyzing heat transfer in these complex nuclear systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106061"},"PeriodicalIF":3.2,"publicationDate":"2025-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145118867","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":"Actinides and depleted uranium as an alternative for sustainable nuclear energy in thermal reactors","authors":"Guilherme Valdir Marchiori da Silva,&nbsp;Lucas Vinicius Gonçalves Chaves,&nbsp;Claubia Pereira","doi":"10.1016/j.pnucene.2025.106057","DOIUrl":"10.1016/j.pnucene.2025.106057","url":null,"abstract":"<div><div>Spent fuel reprocessing can be an alternative to address the environmental and social challenges associated with nuclear energy. This study investigates the use of depleted uranium, stored at the beginning of the cycle, in the spiking with the reprocessed fuel, eliminating the need for mining new fuel or thorium for spiking. Additionally, the radiotoxicity of burned recycled nuclear fuels is analyzed, and their composition is compared with that of thorium and depleted uranium, demonstrating that the use of both results in similar radiotoxicity characteristics. The work also reinforces the use of methods GANEX and UREX + due to the efficiency of proliferation risks compared to PUREX. The results show that the amount of spent fuel in a cycle and the depleted uranium stored in the beginning of the cycle are sufficient to constitute a reprocessed fuel to be inserted in the reactor. There is no big difference between the material required for the proliferative resistance methods and PUREX. The DU-spiked reprocessed fuels behave more similarly to standard UOx regarding radiotoxicity and decay heat, whereas Th-spiked fuels exhibit higher long-term radiological impact due to U-233 production. Depleted uranium in reprocessing offers environmental and waste management advantages, while proliferation resistant methods such as GANEX and UREX + enhance safety and societal acceptance.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106057"},"PeriodicalIF":3.2,"publicationDate":"2025-09-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145118865","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":"Particle swarm optimization of a Small modular reactor fuel assembly design with composed of U-ZR plates and MOX pins, based on Radkowsky concept","authors":"Giovanni Laranjo de Stefani,&nbsp;Thiago Augusto dos Santos,&nbsp;Arthur Leal de Albuquerque,&nbsp;Roberto Junior Artuso Miranda,&nbsp;Marcelo Vilela da Silva,&nbsp;José Rafael Nicolao Carneiro,&nbsp;Maria Vitória de Aguiar Oliveira,&nbsp;Roberto Schirru","doi":"10.1016/j.pnucene.2025.106058","DOIUrl":"10.1016/j.pnucene.2025.106058","url":null,"abstract":"<div><div>This work proposes a fuel assembly design based on the Seed-Blanket Unit (SBU) concept, characterized by two distinct regions: Uranium plates in the central region (seed) to enhance neutron leakage and the surrounding region (blanket) consists of mixed oxide fuel rods composed of thorium and uranium, as originally proposed by Radkowsky. This configuration leverages the advantages of enhanced safety, improved efficiency, and greater resistance to nuclear proliferation. An innovation introduced in this study is the integration of two different geometries within a single fuel assembly: plate-type elements in the seed region—favoring increased neutron leakage—and cylindrical rods in the blanket region—minimizing neutron leakage. This hybrid geometry aims to optimize the neutron economy by enhancing fissile material breeding while controlling neutron losses. The main objective of this research is to investigate the viability and neutronic performance of a novel multigeometry fuel assembly—featuring a thorium-based blanket and a metallic uranium seed—by using Particle Swarm Optimization (PSO) as a computational tool to determine optimal design parameters. The goal is to maximize the in-core breeding of U-233, thereby extending fuel cycle longevity, reducing reliance on enriched uranium, minimizing plutonium generation, and lowering the radiotoxicity and decay heat of spent fuel. Neutronic simulations were carried out using the SERPENT 2.1.30 Monte Carlo code. To couple SERPENT with the PSO algorithm, a Python-based interface was developed to automatically update input parameters and generate simulation files within predefined value ranges. The optimization process was successful, yielding a fuel assembly with a 25.40 % higher conversion factor compared to a reference configuration at the end of the burnup cycle. Additionally, the optimized assembly achieved the production of 2.08 kg of U-233 over a burnup period exceeding 550 effective full-power days (EFPDs).</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106058"},"PeriodicalIF":3.2,"publicationDate":"2025-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099327","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":"Exploring diversion-pathway analysis of a generic molten-salt fast reactor using multiphysics informed signatures","authors":"Samuel Walker ,&nbsp;Ryan Stewart ,&nbsp;Odera Dim ,&nbsp;Emerald Ryan ,&nbsp;Carlos Soto ,&nbsp;Rodrigo de Oliveira ,&nbsp;Parikshit Bajpai ,&nbsp;Cihang Lu","doi":"10.1016/j.pnucene.2025.106047","DOIUrl":"10.1016/j.pnucene.2025.106047","url":null,"abstract":"<div><div>Multiple commercial ventures are exploring molten-salt reactors due to their inherent safety features, flexibility in fuel sources, high fuel utilization, and thermal efficiency. The continual flow of fuel salt, large quantities of fissile material, and the ability to add or divert material due to the liquid nature of fuel salts introduces new challenges for international safeguards. To understand how international safeguards could most efficiently and effectively be applied, it is important to capture the inherent multiphysics nature of a molten-salt reactor. This work examines a generic molten-salt fast reactor to understand how potential diversion scenarios would affect the concentration of radionuclides in the primary and auxiliary systems. Three types of diversion were examined: protracted uranium diversion of fuel salt, gaseous plutonium extraction, and uranium-metal plating. Five isotopes (with corresponding gamma signatures) were found to be statistically different between the various diversion cases and nominal operations. These isotopes are present during the diversion scenarios and continue to be present during operations afterwards indicating that diversion could be detected throughout operations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106047"},"PeriodicalIF":3.2,"publicationDate":"2025-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099380","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":"Recent development and validation of thermal-hydraulic simulation codes for steam generators based on porous media approach","authors":"Yoshiteru Komuro ,&nbsp;Takashi Hibiki","doi":"10.1016/j.pnucene.2025.106050","DOIUrl":"10.1016/j.pnucene.2025.106050","url":null,"abstract":"<div><div>A steam generator functions as a heat exchanger within nuclear power facilities, transferring heat from the primary coolant to the secondary coolant, which then produces steam using thermal energy from the reactor core. When designing steam generators, it is crucial to assess the vibrations of tube bundles caused by the gas-liquid two-phase flow occurring outside the heat transfer tubes. Numerous 3-D thermal-hydraulic codes have been developed and validated over the years. This paper offers a thorough review of recent 3-D thermal-hydraulic codes for steam generators, focusing on publications from 2014 to 2023. It summarizes the constitutive equations related to heat transfer and flow resistance included in these recent codes, providing an overview of the latest understanding of gas-liquid two-phase flow. Furthermore, the paper analyzes and presents experimental data used for code validation, along with the validation results of these recent codes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"191 ","pages":"Article 106050"},"PeriodicalIF":3.2,"publicationDate":"2025-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145099373","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}