Nuclear Engineering and Design最新文献

{"title":"Performance analysis of Dowtherm A heat pipe with internal vapor monitoring","authors":"Mitchell Stephenson ,&nbsp;Trevor Melsheimer ,&nbsp;Joseph Seo ,&nbsp;Abdulbasit Aloufi ,&nbsp;Hansol Kim ,&nbsp;Yassin A. Hassan","doi":"10.1016/j.nucengdes.2025.114049","DOIUrl":"10.1016/j.nucengdes.2025.114049","url":null,"abstract":"<div><div>Medium-temperature heat pipes, operating in the 200–600 °C range, find widespread application in sectors such as nuclear microreactors, solar energy collectors, thermal energy storage, and space. Efficient, passive heat transfer devices, like heat pipes, are essential for power systems operating in this temperature range. Despite such a broad range, traditional working fluids for heat pipes in the medium-temperature regime frequently underperform, prompting the need for more research into these working fluids. Dowtherm A is attractive for its chemical compatibility with heat pipe materials, low toxicity, low flammability, and adequate thermal–hydraulic properties, things that cannot be said for most medium-temperature heat pipe working fluids. This experimental study investigates the performance of Dowtherm A as a medium-temperature heat pipe working fluid, using internal and external measurements to quantify the heat transport in the heat pipe. A 25.4 mm outer diameter, 316 stainless steel tube was used for the heat pipe testing. Ten wraps of 100 × 100 (100 openings per inch) 316 stainless steel screen mesh were used as the wick, with a sliding fit and no annular gap. A fill ratio of 103 % of the total wick void volume was used. An air jacket was attached to the condenser of the heat pipe for cooling. Internal and external temperature measurement was performed, utilizing optical fiber distributed temperature sensing and conventional thermocouples, respectively. All tests conducted were in the horizontal orientation. The test matrix consisted of three different cooling conditions, controlled by changing the flow rate of air in the jacket over the condenser, with multiple power levels for each cooling condition. It was found that the thermal resistance of the heat pipe is not influenced directly by the cooling flow rate but is instead linked to the operating temperature. A minimum thermal resistance of 0.58 °C/W was achieved at the highest operating temperature tested of 274 °C. This corresponds to a maximum effective thermal conductivity of 2300  W/m·K. This finding agrees with values from previous studies. Internal vapor temperature measurements determined the active condenser length, where vapor condenses—a useful tool in heat pipe design. The capillary limit, which governs power transport in heat pipes, was exceeded in all tests without dryout, suggesting Dowtherm A outperformed expectations. This finding questions the soundness of the commonly used theoretical capillary limit, as applied for organic fluids such as Dowtherm A. Collectively, these findings highlight Dowtherm A’s viability for use in medium-temperature heat pipes, offering improved efficiency and operational safety in diverse energy systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114049"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817334","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":"Assessment of a novel fuel block and core arrangement for use in a nitrogen-cooled direct cycle high temperature gas-cooled reactor","authors":"Jeremy Henry Owston","doi":"10.1016/j.nucengdes.2025.114040","DOIUrl":"10.1016/j.nucengdes.2025.114040","url":null,"abstract":"<div><div>This paper investigates the thermal performance of a new fuel block design utilising annular compacts supported in counterbored holes through a coupled thermohydraulic and neutron transport study of a prospective High Temperature Gas cooled Reactor (HTGR) core. The paper highlights the design freedom afforded by a fuel block design which permits irregular spacings of fuel channels without impacting the heat transport to the coolant channel. The approach of optimising the moderating ratio through variable fuel spacings can flatten the radial thermal flux profile, achieving thermal peaking flux factors of less than 1.15 for cores studied in this paper. Flattening the radial thermal flux is shown to minimise variations in coolant outlet temperatures and therefore significantly reduce peak fuel temperatures in the core.</div><div>Burn-up studies of the core demonstrate the benefits of a radially optimised thermal flux profile by demonstrating insensitivity to the fuel burnup of the power profile within the core. This insensitivity results in consistent peak coolant outlet temperatures and small variations in peak fuel temperature over the course of the core life.</div><div>The paper also demonstrates the design flexibility offered by using variable diameter coolant channel displacer rods within the centre of each fuel channel to enhance heat transfer, whilst also balancing the flow distribution within the core. Specifically, the approach of utilising variable displacer rod diameters to match local coolant mass flow with fuel column power is shown to reduce peak fuel temperatures where significant power peaking factors exist.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114040"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799001","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":"Thermo-economic-environmental analysis and performance-based Pareto optimization of a floating nuclear power plant","authors":"Masoud Nasouri ,&nbsp;Navid Delgarm","doi":"10.1016/j.nucengdes.2025.114013","DOIUrl":"10.1016/j.nucengdes.2025.114013","url":null,"abstract":"&lt;div&gt;&lt;div&gt;This paper presents a comprehensive thermo-economic-environmental analysis and performance-based Pareto optimization of a Floating Small Modular Reactor Power Plant (F-SMRP) along the Bushehr coast (Iran), designed to meet regional electricity demands. Innovative ideas are employed to predict the thermodynamic properties of the F-SMRP using an artificial neural network. Upon model verification, a detailed 4E (energy, exergy, exergoeconomic, and environmental economics) analysis is conducted. Further, performance optimization is carried out targeting key metrics such as exergy efficiency (&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;II&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;), the total capital cost rate (&lt;span&gt;&lt;math&gt;&lt;msubsup&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;tot&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;/mrow&gt;&lt;/msubsup&gt;&lt;/math&gt;&lt;/span&gt;), and the total product exergy cost rate (&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;) using the artificial bee colony algorithm. The final optimal configuration, referred to as the Optimized F-SMRP (OF-SMRP), is determined through the analytic hierarchy process decision-making. The results demonstrate that F-SMRP achieves &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;II&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; of 29.9 % and 64.12 %, respectively, with corresponding &lt;span&gt;&lt;math&gt;&lt;msubsup&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;tot&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;/mrow&gt;&lt;/msubsup&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; of $331.8/hour and $6191.8/hour. In contrast, OF-SMRP exhibits notable improvements across all metrics compared to the F-SMRP. The &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; reaches 33.2 %, showing a significant increase of 11 %. Also, &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mover&gt;&lt;mi&gt;W&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mi&gt;e&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; is 34.81 MW, showcasing an increase of 2.11 MW. Similarly, the &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;II&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; improves to 68.1 %, representing a 3.97 % gain. Despite a moderate rise in &lt;span&gt;&lt;math&gt;&lt;msubsup&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;tot&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mspace&gt;&lt;/mspace&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;/mrow&gt;&lt;/msubsup","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114013"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791708","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":"Criticality and sensitivity analysis of VVER-1000 mock-up with SCALE and MCNP5 code using ENDF/B-VII.1 nuclear data library","authors":"Mohammad Omar Faruk ,&nbsp;Mohammad Abdul Motalab ,&nbsp;Mohammad Sayem Mahmood ,&nbsp;Gil Soo Lee","doi":"10.1016/j.nucengdes.2025.114015","DOIUrl":"10.1016/j.nucengdes.2025.114015","url":null,"abstract":"<div><div>Accurate analysis of reactor criticality is essential for reactor design and safety assessments. This paper conducts a criticality study of a VVER-1000 mock-up benchmark experiment, which was performed at the LR-0 research reactor operated by the Research Center Rez in the Czech Republic. Benchmark calculations are performed using two Monte Carlo codes – SCALE (KENO-VI) and MCNP5 – utilizing the ENDF/B-VII.1 continuous-energy nuclear data library for criticality calculations. The mock-up was examined under six different critical configurations by varying coolant levels and boric acid concentrations. This paper provides a comparative analysis of the results from SCALE (KENO-VI) and MCNP5 to assess the suitability of SCALE (KENO-VI) as a verification tool in the regulatory process, with MCNP5 as the reference code. Additionally, the research work also investigates the sensitivity of various reactor system parameters’ uncertainty, highlighting their significant impact on criticality result, which could potentially lead to overly conservative safety margin. The study focuses uncertainty on five key technological parameters: fuel assembly pitch, fuel cladding thickness, fuel density, fuel enrichment and boric acid concentration. A comprehensive analysis of these uncertainties, along with an assessment of their sensitivity to the criticality results, is provided in this study.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791133","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":"Investigating the nexus between radiolysis using spent nuclear fuel and hydrogen production, with environmental safety considerations – A literature review","authors":"Ephraim Bonah Agyekum ,&nbsp;Flavio Odoi-Yorke ,&nbsp;Mustafa Abdullah ,&nbsp;Prangon Chowdhury","doi":"10.1016/j.nucengdes.2025.114048","DOIUrl":"10.1016/j.nucengdes.2025.114048","url":null,"abstract":"<div><div>Given the detrimental consequences of climate change on the environment, hydrogen appears to be one of the solutions to possible decarbonization. Although it is a potential option, the subject of hydrogen production utilizing nuclear spent fuel (SNF) through water radiolysis has not received much attention to explore its potential use for large-scale hydrogen production. This review, therefore, presents an overview of research work on the use of SNF’s ionizing radiation for the production of hydrogen and its potential impact on environmental safety. This paper investigates the advantages and difficulties of hydrogen generation in SNF storage system by utilizing a bibliometric and systematic review technique to analyse previous research works. The research themes were classified into motor, niche, emerging/declining, and basic themes. Some important themes that were found to be central to the topic included radiation shielding, hydrogen production, and environmental sustainability; life cycle assessment; renewable energy integration; nuclear waste management; hydrogen storage; and modular reactors. The study identified some potential research gaps and provided some recommendations for future research. This includes the improvement in the hydrogen detection systems, hydrogen collection and ventilation, containment of radioactive isotopes, understanding radiolysis impacts, and the development of purification and storage methods for hydrogen from spent nuclear fuel. The outcome of this study is expected to shape future research on the subject matter and serve as the foundation for deeper understanding of radiolysis for commercial-scale hydrogen production.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114048"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791716","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 analysis of a water-LBE interaction experiment: Sensitivity analysis, inverse uncertainty quantification and uncertainty propagation","authors":"Qing Zhou,&nbsp;Xin’an Wang,&nbsp;Feng Mao","doi":"10.1016/j.nucengdes.2025.114035","DOIUrl":"10.1016/j.nucengdes.2025.114035","url":null,"abstract":"<div><div>Accurately predicting reactor behavior during Steam Generator Tube Rupture (SGTR) events in Lead-Bismuth Cooled Fast Reactors (LBFRs) is critical for ensuring safety and reliability. This study employs a two-dimensional numerical simulation model to analyze water-LBE interactions during SGTR events with an experiment conducted at the China Nuclear Power Technology Research Institute (CNPRI). A comprehensive sensitivity analysis identified key input parameters that significantly influence pressure responses within the primary pool. Utilizing the Input Parameter Range Evaluation Methodology (IPREM), uncertainty ranges for these parameters were systematically determined. Monte Carlo method, guided by Wilks’ formula, propagated these uncertainties to generate uncertainty bands for pressure responses. The simulation results demonstrated that the uncertainty bands effectively encompassed the observed pressure transients, confirming the model’s reliability.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114035"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799002","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":"Bottom-up levelized cost estimation of low-enriched and low-pressure nuclear batteries","authors":"Gyutae Park,&nbsp;Jacopo Buongiorno,&nbsp;Koroush Shirvan","doi":"10.1016/j.nucengdes.2025.113936","DOIUrl":"10.1016/j.nucengdes.2025.113936","url":null,"abstract":"<div><div>Nuclear batteries (NBs), also known as microreactors, can potentially reduce development and deployment timeline for nuclear energy. However, their lack of economy-of-scale challenges their ability to achieve reasonable cost of energy production. We developed a cost analysis tool that can guide designers to the key attributes that enable cost reduction for NBs. These attractive features are embedded into two NBs for the near-term markets (low enriched uranium fuel, proven experience, and well-known core materials)—the sodium-cooled graphite moderated thermal reactor and the organic-cooled water-moderated thermal reactor. Individual reactor concepts and their point design are presented. Both systems operate at low pressure, which further simplifies the design and operation. The levelized cost of each NB plant operating as a single-unit plant at 15 MWth, are compared for a successful first-of-a-kind production (FOAK) unit, and the Nth-of-a-kind (NOAK) unit. Then, three NB cost-reduction schemes are explored: power uprates, co-siting and equipment sharing, and multi-batch fueling. Based on the levelized costs, the 1-unit, sodium-cooled and organic-cooled NBs for both FOAK and NOAK units would be competitive only in remote markets. However, through a combination of the three NB cost-reduction strategies, the organic-cooled NB would become competitive also in larger markets in the U.S. We find that the most effective parameters in the order of reducing NB’s costs are: 1) higher power reactor designs, 2) multiple-reactor-unit-plants, and 3) batched-fueling scheme.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 113936"},"PeriodicalIF":1.9,"publicationDate":"2025-04-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785934","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":"Conceptual core design study of a pipe type transportable molten salt fast reactor","authors":"Seong Jun Yoon,&nbsp;Tongkyu Park,&nbsp;Sung Kyun Zee,&nbsp;Jae Uk Seo,&nbsp;Yubin Go","doi":"10.1016/j.nucengdes.2025.114033","DOIUrl":"10.1016/j.nucengdes.2025.114033","url":null,"abstract":"<div><div>This study presents a new design for a transportable micro molten salt reactor (MSR) that diverges from conventional upright cylindrical configurations by utilizing a horizontally elongated, thin-walled pipe structure. This novel design aims to facilitate secure transportation and enhance the formation of a molten salt flow field within the reactor core This innovative design aims to facilitate secure transport and enhance the formation of a molten salt flow field within the reactor core. Emphasizing compactness for container loading, the reactor maximizes reflector efficiency and integrates adaptable control mechanisms suitable for its configuration. The horizontally elongated pipe reactor concept allows for the optimal arrangement of subsystems, enhancing vibration safety during transportation by lowering the overall system’s center of gravity, which, in turn, improves durability against vibrations and external impacts.</div><div>The reactor’s total dimensions are 194.68 cm in width, 185.84 cm in length, and 133.84 cm in height, incorporating U-shaped geometries with a 40 cm diameter. The single reactor system meets the target reactivity of 1.03 or higher at the beginning of the cycle and is capable of continuous operation at a 10 MWth output for a period of 3 years. By employing linear and U-shaped geometries, this design reduces the overall thickness and length while offering the flexibility to extend the reactor’s length to meet varying output requirements. This work highlights the potential of a transportable and efficient micro MSR to meet the growing demand for distributed sustainable energy solutions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114033"},"PeriodicalIF":1.9,"publicationDate":"2025-04-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783870","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":"Modelling FeCrAl cladding thermo-mechanical performance: CIEMAT’s contribution to IAEA/CRP ATF-TS","authors":"Pau Aragón,&nbsp;Francisco Feria,&nbsp;Luis E. Herranz","doi":"10.1016/j.nucengdes.2025.114034","DOIUrl":"10.1016/j.nucengdes.2025.114034","url":null,"abstract":"<div><div>This paper provides insights into the response of the advanced technology fuel (ATF) cladding FeCrAl during postulated design basis accident (DBA) and design extension condition without significant fuel degradation (DEC-A) scenarios. Such insights are gained through the development and application of in-house extensions of the FRAPCON/FRAPTRAN fuel performance codes, coupled with the statistical tool DAKOTA, within the framework of a loss-of-coolant accident (LOCA) safety evaluation methodology. While most of the specific FeCrAl models and correlations embedded in these extensions have been documented in the existing literature, the derivation of an instantaneous plasticity model describing the strain-hardening behaviour of FeCrAl alloy C26M is presented for the first time in this paper. The application of the methodology to the DEC-A/LOCA scenario suggests an improved performance of the advanced cladding material, as it maintains its integrity, in contrast to Zircaloy. However, in the DBA/LOCA scenario, no significant differences between these cladding materials were observed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114034"},"PeriodicalIF":1.9,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777470","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":"Aircraft impact: Coupled dynamic simulations part 1: Modelling aspects","authors":"Pierre Wörndle ,&nbsp;Robert Borsutzky ,&nbsp;Thomas Schubert","doi":"10.1016/j.nucengdes.2025.114046","DOIUrl":"10.1016/j.nucengdes.2025.114046","url":null,"abstract":"<div><div>During the last five decades, the load case “aircraft impact” changed constantly with regard to the definition of the threat, the applied analysis methods, and the required checks; particularly in the analysis of concrete structures in the nuclear sector. On the one hand, the load case was extended from military aircrafts to commercial aircrafts and on the other hand, the rapid development of computer hardware allowed more and more complex numerical simulations with detailed modelling of aircraft and building structure in a coupled analysis. Keeping this development in mind, as well as additional confidential aspects, increasing authorities’ requirements and the lack of detailed normative guidelines, these kinds of specialized analyses can only be performed with deep knowledge not only of the general aspects of the aircraft impact load cases but also of the complex numerical modelling and simulation of these situations.</div><div>This article is the first part of a two-part publication. In terms of content, the two publications focus on the two overarching topics that need to be considered in a typical aircraft impact assessment: on the one hand, the aspects of modeling and structural analysis and, on the other, the assessment of vibrations and the design of mitigation measures. The first part of this paper covers critical modeling aspects within the scope of a coupled structural analysis as well as aspects of assessing the structural damage. For this, this paper summarizes some of the main issues regarding the actual state-of-the-art Aircraft Impact Analysis (AIA) and points out the most discussed topics with focus on numerical modelling aspects in coupled dynamic AIA simulations.</div><div>One of the main conclusions for a state-of-the-art Aircraft Impact Analysis (AIA) is the recommendation of a detailed numerical aircraft model or at least of the center wing box and the attached wings and a coupled dynamic simulation with the impacted structure. With regard to the introduced shear forces, this approach is the only one that allows for the necessary detailed analysis of local impact effects and ultimately leads to a comprehensive assessment of the structural resistance.</div><div>In the second part of this publication, Wörndle and Borsutzky (planned to be published), the aspects of assessment criteria regarding induced vibrations are discussed, supplemented by examples of mitigation methods.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114046"},"PeriodicalIF":1.9,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777469","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}