Nuclear Engineering and Design最新文献

{"title":"Sensitivity study of hydrogen Soret transport in yttrium Hydride-Based nuclear fuel","authors":"Jordan A. Evans ,&nbsp;Chase N. Taylor ,&nbsp;Adrian R. Wagner ,&nbsp;Ryan T. Sweet ,&nbsp;Travis L. Lange ,&nbsp;Nicolas E. Woolstenhulme","doi":"10.1016/j.nucengdes.2025.114030","DOIUrl":"10.1016/j.nucengdes.2025.114030","url":null,"abstract":"<div><div>Yttrium hydride is an excellent solid neutron moderator material for high temperature nuclear reactor applications due to its high hydrogen density and exceptional hydride stability at high temperatures. Despite these attractive characteristics, the details of how hydrogen behaves within yttrium hydride while temperature gradients exist are still not well understood. The evolution of the hydrogen composition profile resulting from a temperature gradient requires knowledge of hydrogen’s heat of transport, a critical parameter that has not yet been measured for this material. In this work, we perform hydride redistribution, hydrogen dissociation, and hydrogen leakage calculations while varying the Soret heat of transport of hydrogen in yttrium hydride to elucidate the sensitivity of hydride stability under temperature gradients to this parameter. This study analyzes hydride stability of a hypothetical uranium-yttrium hydride nuclear fuel design during operation of a high temperature liquid metal-cooled nuclear reactor. Assuming U-YH_x could be fabricated in a physically stabilized manner, this fuel system can likely maintain hydride stability while operating at very high power densities and temperatures. We find that even though the hydrogen dissociation pressure in the gas gap does vary by several percent as the heat of transport temperature parameter is varied, the hydrogen content in the U-YH_x fuel meat is relatively insensitive to this parameter over the course of a high burnup fuel cycle; this is due to yttrium hydride’s excellent hydrogen retention under the high temperature conditions considered here. This suggests that hydride stability analyses are insensitive to the value of the Soret heat of transport in U-YH_x under steady state liquid metal-cooled reactor conditions. However, the susceptibility to internal gas overpressurization-induced stress-rupture of the cladding during a high temperature transient is more sensitive to this parameter due to the non-linear dependence of hydrogen gas dissociation pressure vs. composition and temperature.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114030"},"PeriodicalIF":1.9,"publicationDate":"2025-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143824333","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":"Multiphysics simulation and analysis for fuel behavior with geometric irregularities of missing pellet surface and eccentricity","authors":"Xiaoyang Yuan,&nbsp;Rong Liu,&nbsp;Shengyu Liu","doi":"10.1016/j.nucengdes.2025.114039","DOIUrl":"10.1016/j.nucengdes.2025.114039","url":null,"abstract":"<div><div>Missing pellet surface (MPS) defect and fuel eccentricity are both the abnormal geometric phenomena of nuclear fuel rods. One of cladding failure causes is ascribed to the MPS owing to manufacturing, and fuel eccentricity will lead to irregular temperature distribution which could affect the design and safe operation of nuclear reactor. However, most of nuclear fuel performance codes are developed with 1.5D and 2D axisymmetric geometries and not applicable for these asymmetric problems. In this paper, a code using 3D geometric model is established to simulate fuel pellet with irregular geometries of MPS and eccentricity based on COMSOL Multiphysics software. First, the existence of MPS is considered and analyzed. The simulation results of MPS defect in conditions of stable power, power change and reactivity initiated accident (RIA) condition are discussed, and some adverse effects on thermal and mechanical performance can be observed under these conditions. The discussion of depth variation of MPS and different fuel types is also included. Finally, the effects of eccentricity on fuel behavior in different cases are researched. Fuel eccentricity can lead to uneven temperature field and early pellet-cladding mechanical interaction (PCMI) time.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114039"},"PeriodicalIF":1.9,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143820501","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":"Optimizing Micro-PeLUIt reactor with UO2-ThO2 fuel mixtures and improved graphite moderation","authors":"Ahmad Muzaki Mabruri ,&nbsp;Nuri Trianti ,&nbsp;Zaki Su’ud ,&nbsp;Ratna Dewi Syarifah","doi":"10.1016/j.nucengdes.2025.114051","DOIUrl":"10.1016/j.nucengdes.2025.114051","url":null,"abstract":"<div><div>Micro-PeLUIt is a High-Temperature Reactor (HTR) design similar to China’s HTR-10 reactor, developed by Indonesia to meet commercial and industrial power demands with flexible operational power ranging from 10 MWt to 40 MWt. The latest Micro-PeLUIt pebble fuel design is proposed to feature lower ²³⁵U enrichment levels and higher heavy metal (HM) content per pebble compared to the standard HTR-10 fuel. These conditions may pose challenges regarding the criticality lifetime of the fuel, particularly due to reduced moderation effects. Another issue with this design is the potential increase in plutonium production, which raises concerns about fuel waste management and nuclear proliferation. This study proposes the use of ²³²Th as a UO₂-ThO₂ fuel mixture to reduce neutron absorption by ²³⁸U, thereby limiting plutonium production. Two UO₂-ThO₂ mixing methods are evaluated: mixing in a single TRISO kernel as compound (CM) and mixing in different TRISO kernels (TM). The optimal UO₂-ThO₂ design is also evaluated with adjusted ²³⁵U enrichment specifications for Micro-PeLUIt, as well as graphite density adjustments in the pebble matrix to enhance neutron moderation. The results show six optimal UO₂-ThO₂ fuel variations capable of reducing plutonium production by 10%–25% and decreasing total fissile material per pebble by 9%–26%. The use of UO₂-ThO₂ mixed fuel with graphite density adjustment in the pebble matrix can result in performance similar to normal fuel in the HTR-10 reactor. Increased graphite density enhances neutron moderation within the pebble, effectively maintaining the criticality of the system without significantly increasing the HM loading per pebble. Furthermore, this design provides better criticality potential compared to UO₂ fuel, making it more efficient in operation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114051"},"PeriodicalIF":1.9,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817385","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":"Research on a rapid prediction method for multi-physics coupled fields in small lead-cooled fast reactors based on machine learning","authors":"Haochen Huang ,&nbsp;Daogang Lu ,&nbsp;Yu Liu ,&nbsp;Danting Sui ,&nbsp;Fei Xie ,&nbsp;Hao Ding","doi":"10.1016/j.nucengdes.2025.114065","DOIUrl":"10.1016/j.nucengdes.2025.114065","url":null,"abstract":"<div><div>The small lead-cooled fast reactor (LFR), as a typical representative of the fourth-generation reactors, is widely applicable to island areas, deep-sea environments, and specialized industrial scenarios due to its high power density, modular design, and inherent safety features. The neutron physics, thermal-hydraulics, and structural deformation in LFR are highly coupled, and most existing studies neglect the impact of fuel deformation on the neutron physics and thermal-hydraulics, making it difficult to accurately reflect the operating state in the reactor. However, the coupling analysis of the three fields involves significant computational costs, making it challenging to achieve real-time prediction. To address this issue, this paper proposes a method that integrates the multi-physics field coupling of reactors with machine learning-based rapid prediction techniques. By using measurable parameters of the reactor, rapid prediction of the multi-physics field distribution inside the core can be achieved. The final test results show that the model controls the relative prediction error of each physical field within 1%, with prediction time significantly shortened compared to traditional numerical methods. It efficiently achieves accurate and rapid prediction of the multi-physics field distribution in the LFR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114065"},"PeriodicalIF":1.9,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817338","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":"Experimental and numerical analysis on safety condenser performance based on P1 experiments at the PKL facility","authors":"I. Gómez-García-Toraño ,&nbsp;S. Schollenberger ,&nbsp;L. Dennhardt ,&nbsp;A. Wielenberg ,&nbsp;M. Vernassière ,&nbsp;S. Buchholz ,&nbsp;O.S. Al-Yahia ,&nbsp;E. Garcia ,&nbsp;M. Polidori ,&nbsp;N. Sobecki ,&nbsp;F. Lahovský ,&nbsp;F. de-Bouet-du-Portal ,&nbsp;G. Grippo ,&nbsp;M. Montout","doi":"10.1016/j.nucengdes.2025.114016","DOIUrl":"10.1016/j.nucengdes.2025.114016","url":null,"abstract":"<div><div>Passive systems are being considered for advanced reactor designs because of their enhanced reliability against an extended loss of offsite power. Particularly, the SAfety COndenser (SACO) stands out because of its capacity of passively removing core decay heat through the steam generators by condensing steam inside a immersed heat exchanger. This article presents recent experimental data and the associated numerical calculations on the vertical straight-tube SACO installed at the PKL facility. In particular, the SACO power removal capacity has been studied within the frame of test P1.1 consisting of steady state phases A, B, C and D with varying pool liquid levels and a Core Exit Temperature of 237 °C i.e. 20 K subcooling.</div><div>Experimental results show the SACO capability to transfer its nominal power of 450 kW despite the accumulation of nitrogen in the straight tubes. Improved venting procedures of phases A2 and C2 allowed a partial removal of nitrogen from the tubes and hence, an increase of the maximum core power to keep the CET constant in comparison to their counterpart phases A and C. The accumulation of nitrogen in the tubes leads to the formation of passive zones characterised by a degraded heat transfer towards the pool and significant cool-down of the liquid film.</div><div>An important numerical work has also been conducted using the CATHARE-3, ATHLET, TRACE, RELAP-5 system thermalhydraulic codes and <span><math><mtext>Neptune_cfd</mtext></math></span>, either in standalone mode or coupled with CATHARE-2. Several approaches have been adopted in order to model the primary system, SACO pool, straight tubes, boundary and initial conditions (e.g. nitrogen content, heat losses), auxiliary components (heaters, pump cooling), which add up to the physical models when analysing discrepancies with experimental results. Generally, codes are able to predict the phenomena happening in PKL, although further efforts should be invested in the use of 3D approaches to model the pool and the improvement of condensation modelling in vertical tubes for the SACO-operating region.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114016"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817333","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":"Blast response of a scaled reinforced concrete structure with Two-Leaf cavity infill wall","authors":"Ahmet Tuğrul Toy ,&nbsp;Onur Onat ,&nbsp;Barış Sevim","doi":"10.1016/j.nucengdes.2025.114055","DOIUrl":"10.1016/j.nucengdes.2025.114055","url":null,"abstract":"<div><div>Nuclear power plants, composed of boiler houses, reactors, and other facilities, operate at a high risk of explosion. Engineers design boiler hoses and other facilities to withstand dynamic loads like earthquakes, machine vibrations, wind, and blast loads. However, over time, these structures may cease to meet the requirements of current codes. Therefore, it remains unclear how different materials, their orientations, and their interactions, such as masonry and reinforced concrete, will respond in the event of a blast around a nuclear power plant. Currently, this study aims to evaluate the global and local blast response of single and two-leaf cavity infill wall enclosures with reinforced concrete structures. For this purpose, a scaled structure that is exposed to a shake table experiment has been selected. Then the structural system is numerically modelled by using ANSYS-AUTODYN and calibrated based on dynamic identification tests. The explosive amount is fixed at 78 kg to facilitate comparison of two models. For blast analysis of the structural system, two different infill wall typologies and three different scenarios are evaluated. The location of explosives determined the studied cases. We register the analytical blast responses in terms of the pressure, strain, and out-of-plane displacement of the infill wall. We limited the blast analyses to 3 ms. We compared the out-of-plane displacement of single and cavity infill walls with each other and with UFC 3–340-02. According to the findings, the thinner leaf in the Two Leaf Cavity Wall model protects the thicker leaf from damage.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114055"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143807823","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":"Molecular dynamics study of local structure and migration properties of LiCl-Li2O-Li molten salts based on machine-learned deep potential","authors":"Jitang Xu ,&nbsp;Yilin Wang ,&nbsp;Benlin Yao ,&nbsp;Yanhong Jia ,&nbsp;Yiqun Xiao ,&nbsp;Lin Zhang ,&nbsp;Bin Li ,&nbsp;Hui He ,&nbsp;Baohua Yue ,&nbsp;Liuming Yan","doi":"10.1016/j.nucengdes.2025.114052","DOIUrl":"10.1016/j.nucengdes.2025.114052","url":null,"abstract":"<div><div>The local structure and physical properties of LiCl-Li₂O-Li molten salt, the reaction medium for lithium thermal and electrolytic reduction, are very important for the study of spent fuel pyroprocessing process. In this work, the machine-learned deep potential (MLDP) was trained using dataset based on first-principle molecular dynamics (FPMD) and was used to predict the changes in the physical properties of molten LiCl with the addition of different concentrations of Li₂O and Li between 923 K and 1323 K. Deep potential molecular dynamics (DPMD) calculations were performed for properties including shear viscosity, electrical conductivity, thermal conductivity, and specific heat capacity. It was revealed that the addition of Li significantly reduces the diffusion activation energies (<span><math><mrow><msub><mi>E</mi><mi>a</mi></msub></mrow></math></span>) of Li⁺ and Cl^- in the molten salt. By comparison with the experimental data of pure LiCl, it can be concluded that the MLDP can describe the inter-atomic interactions of molten salt correctly, overcome the problem of missing potential parameters in the classical inter-atomic empirical potentials. Finally, DPMD allows to simulate large systems with comparable accuracy of FPMD, thus provide theoretical guidance for the optimization of the pyroprocessing technology.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114052"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817332","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":"Design of novel refractory equiatomic multi-principal elemental alloys based on Mo-Nb-Ti system for Gen IV reactor applications","authors":"Anilas Karimpilakkal ,&nbsp;Joseph W. Newkirk ,&nbsp;Jason L. Schulthess ,&nbsp;Frank Liou ,&nbsp;Visharad Jalan ,&nbsp;Haiming Wen","doi":"10.1016/j.nucengdes.2025.114050","DOIUrl":"10.1016/j.nucengdes.2025.114050","url":null,"abstract":"<div><div>Excellent irradiation damage resistance demonstrated by multi-principal elemental alloys (MPEAs) has sparked significant interest among researchers, prompting exploration into their vast compositional space, to validate their suitability for nuclear applications. A combined approach of thermodynamic and empirical parameters calculations alongside CALPHAD (CALculation of PHAse Diagrams) for phase formation predictions enable high-throughput material selection for sophisticated applications like nuclear, overcoming laborious and time-consuming experiments. Key thermodynamic and empirical parameters for eight novel equiatomic MPEAs, based on seven low thermal neutron cross section refractory elements, for predicting phase formation were calculated, and equilibrium and non-equilibrium simulations in CALPHAD were employed to comprehensively evaluate the systems. Pseudo binary phase diagram simulations showed that Zr, V or equiatomic CrV additions to the base MoNbTi alloy (MoNbTi-Zr, MoNbTi-V and MoNbTi-CrV alloys) favor the formation of isomorphous body-centered cubic (BCC) phase at high temperatures, while Cr, Al, equiatomic ZrV, or equiatomic CrAl additions (MoNbTi-Cr, MoNbTi-Al, MoNbTi-ZrV or MoNbTi-CrAl alloys) limit the solubility of them. Equilibrium CALPHAD simulations at 750 °C were consistent with XRD results on MoNbTi, MoNbTiZr and MoNbTiCr alloys, and partially for others. Notably, elemental segregation observed in the backscattered electron (BSE) scanning electron microscopy (SEM) images of the alloys was accurately simulated through non-equilibrium Scheil solidification calculations in CALPHAD, further verified by experiments. The precipitation of TiCr₂ Laves phase in Cr containing MoNbTiCr and MoNbTiCrAl was accurately predicted while discrepancies were noted in MoNbTiCrV. The equilibrium simulations also provided insights into phase compositions at specific temperatures offering a pathway for tailoring the desired microstructure and properties of these systems. Empirical parameters calculations successfully predicted random solid solution in the base MoNbTi alloy, and with an exception in MoNbTiV and MoNbTiAl, predicted intermetallic precipitation in the rest, especially, Laves phase precipitation in Cr containing alloys.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114050"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817337","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":"Bamboo-SFuel: A nuclide composition evaluation code for PWR spent fuel in NECP-Bamboo","authors":"Yunzhao Li ,&nbsp;Songzhe Wang ,&nbsp;Yuancheng Zhou ,&nbsp;Senhan Yang ,&nbsp;Ruizhi Shao ,&nbsp;Liangzhi Cao ,&nbsp;Tian Chen ,&nbsp;Miao Yu ,&nbsp;Zeng Shao ,&nbsp;Guoming Liu","doi":"10.1016/j.nucengdes.2025.114047","DOIUrl":"10.1016/j.nucengdes.2025.114047","url":null,"abstract":"<div><div>To accurately obtain the nuclide composition in spent fuel, a nuclide composition calculation program named Bamboo-SFuel has been developed as a new component in the PWR-core analysis software NECP-Bamboo. It can switch between different depletion data libraries, including one with 233 nuclides in the lattice code Bamboo-Lattice, the one with 1547 nuclides, or the one with 3838 nuclides. The shortcomings of existing domestic procedures, such as incomplete nuclide types in the data library and incomplete simulation of the irradiation process, limit the reliability and economy of the safety analysis of spent fuel reprocessing. The nuclide composition was quantitatively verified and analyzed using the measured data of Post-Irradiation Experiments (PIE). The source term calculation accuracy of the program was verified by comparing with the calculation results of the SCALE package, and the impact of different depletion data libraries on the nuclide composition and source term calculation results was also explored. Encouraging conclusions have been drawn from the numerical results. (1) Bamboo-SFuel can accurately analyze the radionuclide composition and radioactive source term of spent fuel assemblies of commercial PWR under different irradiation conditions. (2) The depletion data library with different numbers of nuclides has little influence on the calculation results of important nuclide composition, but has a great influence on the total radioactive source term calculation results. (3) Based on the built-in depletion data library containing 1547 nuclides, this program can simultaneously provide reliable radioactive source analysis of spent fuel and important nuclide composition concerned by radiation safety analysis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114047"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817331","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":"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}