Nuclear Engineering and Design最新文献

{"title":"Sensitivity analysis of main neutronic parameters of standard PWR and WWER-1000 pin cells to resonance self-shielding treatment methods and cross section libraries","authors":"Farrokh Khoshahval","doi":"10.1016/j.nucengdes.2026.114801","DOIUrl":"10.1016/j.nucengdes.2026.114801","url":null,"abstract":"<div><div>To evaluate the neutronic behavior of a fuel pellet versus burnup, it is necessary to select a proper lattice code. The deterministic nuclear codes are superior to probabilistic codes in terms of computational speed. The deterministic codes such as WIMS-D5 and DRAGON codes are dependent on the method used for treatment of resonance self-shielding cross sections and the library of cross sections. This paper focus on two standard PWR and WWER pin cells and evaluate both equivalence in dilution, and multiband methods for resonance self-shielding calculations. In addition, three different neutron library cross sections (WLUP-69, WLUP-172, DRAGLIB-172) are assessed. It is revealed that the deterministic lattice DRAGON code can accurately treat the self-shielding behavior in the PWR and WWER pin cell and one can trust on the generated main neutronic parameters and multi-group homogenized cross-sections of PWR and WWER fuel assembly and/or whole-core calculations. In addition, it is proved that generalized Stamm'ler (SHI: self-shielding module of the Dragon) depends on the type of geometry. For square geometries, for all three different libraries, the best values of multiplication factor are obtained using GSM-NOLJ and the worst results are attributed to the GSM-LJ. Furthermore, to have a better evaluation of the applied self-shielding methods and their accuracy, the main isotopic concentration variation and fuel temperature coefficient versus burnup are also computed, considering different libraries and self-shielding treatments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114801"},"PeriodicalIF":2.1,"publicationDate":"2026-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081261","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 the base isolation design and parameter optimization analysis of friction pendulum bearings for reactor building in swimming pool-type low-temperature heating reactor","authors":"Yingying Gan ,&nbsp;Xiaoying Sun ,&nbsp;Pengxiang Dong ,&nbsp;Ziqiao Liu","doi":"10.1016/j.nucengdes.2025.114739","DOIUrl":"10.1016/j.nucengdes.2025.114739","url":null,"abstract":"<div><div>The swimming pool-type low-temperature heating reactor (SPLTHR) is a single-unit small heating reactor that can serve as an alternative to fossil energy. The peak ground acceleration (PGA) for the safe shut- down earthquake (SSE) at the proposed site is up to 0.5 g in horizontal direction. To ensure seismic safety and improve economic efficiency of the reactor, the Friction Pendulum (FP) bearing is employed for the base isolation design of the reactor building. Firstly, a three-dimensional finite element model (FEM) of the reactor building is established. The layout scheme of the base isolation layer is designed. Subsequently, a parameter optimization analysis about the equivalent radius of curvature and dynamic friction coefficient of the FP bearing is conducted to achieve the optimal isolation performance for the reactor building. Finally, the acceleration response spectrum (ARS) in three directions were compared between the base-isolated system and non- isolated system at the same place. The acceleration reduction rate was defined to quantified the isolation performance. The study results indicate that the base isolation layer using 28 FP bearings with load capacity in axial direction of 15,000 kN and 8 viscous damper can meet the design requirement. The dynamic friction coefficient of the FP bearing has a more significant influence on the isolation performance than the equivalent radius of curvature. In general, a larger equivalent radius of curvature and a smaller dynamic friction coefficient result in better isolation performance. The ARS in horizontal direction of the superstructure in the non-isolated system completely envelops that of the base-isolated system. The seismic response of the base-isolated system shows a substantial reduction in the dominant frequency and a significant decrease in the ARS of the superstructure in horizontal direction. The maximum reduction rates for zero-period acceleration (ZPA) and peak acceleration can reach up to 75.0 % and 85.4 %, respectively, demonstrating excellent isolation performance. Compared to the ARS in vertical direction of the non-isolated system, the base-isolated system has a lower dominant frequency, a leftward shift in peak acceleration response (with lower peak frequency), and an insignificant increase in peak values. It is recommended to focus on the seismic response of key equipment which is sensitive to the vertical frequency ranges of the base-isolated system, and to implement appropriate local vertical isolation measures if necessary.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114739"},"PeriodicalIF":2.1,"publicationDate":"2026-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081255","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 fluid-structure interaction characteristics of transient pressure waves in reactor coolant pump under shaft seizure accident","authors":"Teng Niu ,&nbsp;Yi Bin Li ,&nbsp;Hai Long Yuan ,&nbsp;Xue Zhao ,&nbsp;Kong Sheng Liu","doi":"10.1016/j.nucengdes.2026.114795","DOIUrl":"10.1016/j.nucengdes.2026.114795","url":null,"abstract":"<div><div>This study investigates the fluid-structure interaction (FSI) characteristics of transient pressure waves during a reactor coolant pump (RCP) shaft seizure accident (SSA) through bidirectional FSI numerical simulation of the coolant pipeline. Based on a model of the HPR1000 reactor single-loop system with matched resistance characteristics, the analysis focuses on the RCP flow field pressure, internal pressure fluctuations, and pipeline dynamic response. The results demonstrate that coolant pipeline fluid-structure interaction (CPFSI) significantly alters pressure distributions in RCP flow components. During shaft seizure, CPFSI causes a notable expansion of the low-pressure zone at the impeller inlet and an increase in volute pressure. Immediately after shaft seizure, it induces a widespread pressure decrease at the inlet nozzle, a significant enlargement of the low-pressure region within the guide vane flow passage, and a slight expansion of the low-pressure area at the volute outlet. After shaft seizure ends, CPFSI leads to substantial pressure reductions at the inlet nozzle, impeller inlet, volute annular cavity, and volute outlet, alongside a marked expansion of low-pressure zones at the impeller inlet and IPS and a contraction of the high-pressure zone at the GPS inlet. Throughout the shaft seizure transition process, the transition pipe experiences the most pronounced deformation and fluctuation, followed by the hot leg pipe, with the cold leg pipe showing minimal variation. These structural vibrations intensify pressure fluctuations at RCP monitoring points, leading to either attenuation or amplification of transient pressure wave amplitudes. This research reveals the coupling mechanism between transient pressure waves and pipeline dynamics during an SSA, providing a theoretical basis for accurately assessing RCP operational safety.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114795"},"PeriodicalIF":2.1,"publicationDate":"2026-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081263","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":"Side-by-side high-temperature accident performance of ATF and conventional claddings in the CODEX-ATF experiment","authors":"Nóra Vér ,&nbsp;Róbert Farkas ,&nbsp;Berta Bürger ,&nbsp;Anna Pintér-Csordás ,&nbsp;Tamás Novotny ,&nbsp;Erzsébet Perez-Feró ,&nbsp;Péter Szabó ,&nbsp;Levente Illés ,&nbsp;Zoltán Kovács ,&nbsp;Dávid Cinger ,&nbsp;Martin Ševeček ,&nbsp;Zoltán Hózer","doi":"10.1016/j.nucengdes.2026.114770","DOIUrl":"10.1016/j.nucengdes.2026.114770","url":null,"abstract":"<div><div>The CODEX-ATF integral bundle test was conducted within the framework of the IAEA Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS) project at the CODEX (COre Degradation Experiment) facility in Hungary. The electrically heated seven-rod bundle consisted of four Cr-coated and three uncoated Zr alloy cladding tubes, enabling a direct comparison of their behavior under high-temperature accident conditions. The experiment primarily aimed to investigate fuel failure and degradation mechanisms.</div><div>During the test, several rods exhibited ballooning and burst phenomena. The maximum temperature exceeded 1600 °C. The transient was terminated by a bottom-up water quench. The total hydrogen generation was approximately 3 g, indicating substantial oxidation of the zirconium-based components. Intensive Zr-Cr eutectic interaction was observed in the hottest region of the bundle on the Cr-coated claddings. Post-test examinations revealed pronounced deformation and failure in both coated and uncoated claddings.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114770"},"PeriodicalIF":2.1,"publicationDate":"2026-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081254","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":"Multidimensional optimization of the high-diodicity diaphragm hydrodiode for passive safety systems of nuclear power plants","authors":"Victor Shcherba ,&nbsp;Anatoliy Khait ,&nbsp;Sergey Kaigorodov ,&nbsp;Ksenia Sokirko ,&nbsp;Evgeniy Pavlyuchenko","doi":"10.1016/j.nucengdes.2026.114803","DOIUrl":"10.1016/j.nucengdes.2026.114803","url":null,"abstract":"<div><div>A novel high-efficiency diaphragm hydrodiode (i.e., fluidic diode) for NPP safety circuits is proposed. To achieve maximum diodicity, a multi-parameter optimization of its geometry is performed using a machine-learning-aided surrogate model. Training the surrogate model is performed using the quasi-random sampling, while the exact diodicity values were provided by the CFD simulations based on the Reynolds-Averaged Navier-Stokes equations closed with the <span><math><mi>k</mi><mo>−</mo><mi>ω</mi></math></span> turbulence model. Iterative complementation of the sampling is employed to further increase the surrogate model accuracy. Genetic and Trust-Region optimization algorithms are executed on top of the surrogate model to arrive at the optimal hydrodiode configuration. The maximum diodicity value reported by both CFD and the surrogate model is <span><math><msub><mi>D</mi><mi>CFD</mi></msub><mo>≈</mo><mn>2.74</mn></math></span>, while the experimentally confirmed diodicity of the optimal diode configuration is found to be <span><math><msub><mi>D</mi><mi>exp</mi></msub><mo>=</mo><mn>2.59</mn></math></span>. Such a high diodicity value for the diaphragm hydrodiode is reported for the first time, thus constituting an achievement in the field. The proposed design and optimization methodology open up possibilities for constructing compact and reliable passive components for safety systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114803"},"PeriodicalIF":2.1,"publicationDate":"2026-01-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081253","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":"Computational modeling of graphite degradation in molten salt reactors: Role of infiltration","authors":"Veerappan Prithivirajan ,&nbsp;Benjamin Spencer ,&nbsp;Joseph Bass ,&nbsp;Somayajulu L.N. Dhulipala ,&nbsp;Daniel Schwen ,&nbsp;Mustafa K. Jaradat","doi":"10.1016/j.nucengdes.2026.114796","DOIUrl":"10.1016/j.nucengdes.2026.114796","url":null,"abstract":"<div><div>Molten salt reactors (MSRs) often employ graphite as a moderator and reflector. An important challenge for deploying graphite in these reactors is that, due to limited experimental data, our understanding of graphite’s structural integrity in molten salt environments remains incomplete. This study addresses heat generation from fuel-bearing salt that has infiltrated open pores in the graphite, driven primarily by pressure differentials. This is one of multiple identified physical and chemical mechanisms through which molten salt could potentially degrade graphite. Thermally driven stresses are quantified using the Molten-Salt Reactor Experiment (MSRE) graphite moderator elements as a case study. Finite element simulations predict stress distributions at varying infiltration levels, indicating that thermal stresses increase with higher infiltration. Rare-event simulations using the parallel subset simulation framework identify the combinations and corresponding ranges of input parameters that lead to stresses above a specified threshold. In particular, combinations involving high infiltration amounts, high power density, and low thermal conductivity tend to induce the highest stresses. Under the inputs and assumptions considered in this work, the magnitudes of the thermally driven stresses are quite low, with a very low likelihood of causing failure due to exceeding the graphite’s tensile strength. Additionally, rare-event simulations were performed for two more scenarios: a scaled-up moderator geometry and a localized hotspot in the original geometry. Both cases resulted in increased susceptibility to failure, though not to a detrimental extent. Furthermore, the combined effects of irradiation and infiltration-induced thermal stresses were evaluated. The results showed that thermal stresses from infiltration were negligible compared to those caused by irradiation. The findings of such a study are inherently component-specific, but the methodology presented here could be used for similar assessments of salt-infiltration effects in other graphite components.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114796"},"PeriodicalIF":2.1,"publicationDate":"2026-01-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146057606","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":"Leveraging ENDF data in an enhanced ORIGEN2 library for advanced VVER-1000 fuel management","authors":"Saeedeh Arabzadeh ,&nbsp;Seyed Pezhman Shirmardi ,&nbsp;Nasser Mansour Shariflou","doi":"10.1016/j.nucengdes.2026.114758","DOIUrl":"10.1016/j.nucengdes.2026.114758","url":null,"abstract":"<div><div>Accurate burn up calculations are critical for nuclear reactor design, particularly for determining the nuclear concentrations of fuel isotopes and fission products throughout the reactor cycle. An updated cross-sectional library is essential for effective fuel behavior analysis and management. This study aims to develop a tailored cross-sectional library for the VVER-1000 reactor to enhance the accuracy of burn up calculations using the ORIGEN2 code, leveraging the ENDF reference library. The Monte Carlo N-Particle (MCNPX) code was used to generate the required cross-sectionals, which were then integrated into ORIGEN2 for burn up calculations. The results were compared with those obtained using the existing library. The new library demonstrates moderately improved accuracy and computational efficiency for burn up calculations in the VVER-1000 reactor compared to the previous library.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114758"},"PeriodicalIF":2.1,"publicationDate":"2026-01-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146057607","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":"Multi-Physics Benchmark for a Thermal Molten Salt Reactor","authors":"P. Pfahl ,&nbsp;B. Kędzierska ,&nbsp;I. Kim ,&nbsp;A. Chambon ,&nbsp;L. Fischer ,&nbsp;L. Bureš ,&nbsp;J. Groth-Jensen ,&nbsp;Y. Kim ,&nbsp;A. Rineiski ,&nbsp;B. Lauritzen","doi":"10.1016/j.nucengdes.2026.114790","DOIUrl":"10.1016/j.nucengdes.2026.114790","url":null,"abstract":"<div><div>Verification of nuclear codes is an important step in licensing nuclear reactors. For molten salt reactors, the involved physics phenomena are strongly coupled and include those introduced by the movement of liquid fuel that are not present at nominal conditions in solid fuel reactors. This movement of fuel inside and outside the core poses new simulation challenges. In this paper, a benchmark for a graphite-moderated molten salt reactor with a simplified out-of-core model is proposed and studied. The benchmark addresses both neutronics and thermal-hydraulics phenomena, including the delayed neutron precursor drift inside and outside of the active core region, as well as the temperature feedback. As for the thermal-hydraulics, a laminar flow field with conjugate heat transfer, delayed neutron precursor movement, and a simplified heat exchanger is modeled. The benchmark is investigated with the MOOSE tools Griffin and Squirrel, coupled with the MOOSE internal thermal-hydraulics abilities, the Monte Carlo code iMC coupled with OpenFOAM, Nek5000 with a custom point kinetics solver, the coupled neutronics and fluid dynamics code SIMMER with capabilities for severe accident simulations, and the Modelica-based library TRANSFORM. By employing a variety of high- and low-fidelity modeling approaches, a robust comparison across different codes is ensured. OpenMC and Serpent are employed as reference codes to verify the correct implementation of the neutronics. This paper provides a comprehensive comparison of the strengths and weaknesses of the codes and their underlying modeling assumptions. It examines how modeling assumptions affect the steady-state solution and how they propagate into the transient analysis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114790"},"PeriodicalIF":2.1,"publicationDate":"2026-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078926","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":"Prediction of leakage rates from containment buildings under ultimate internal pressure","authors":"Jin-Young Park,&nbsp;Tae-Hyun Kwon","doi":"10.1016/j.nucengdes.2026.114792","DOIUrl":"10.1016/j.nucengdes.2026.114792","url":null,"abstract":"<div><div>Accurate modeling of containment leakage is essential for predicting radioactive release during severe nuclear accidents. Although simplified pressure-based methods provide general estimates, they lack the ability to capture damage progression and determine leakage locations caused by structural failures, such as concrete cracking and liner tearing. This study examines the structural response and leakage behavior of a prestressed concrete containment building representative of the APR-1400 under ultimate internal pressure. A comprehensive three-dimensional nonlinear finite-element (FE) model was developed, and the predicted internal pressures corresponding to characteristic strain levels in major components were consistent with those of the Sandia National Laboratories (SNL) test data and simplified calculations, validating the reliability of the FE model. Four leakage prediction methods were employed to evaluate leakage rates, incorporating both permeation-based and crack development approaches. The analysis reveals that leakage initiates primarily at discontinuities, such as equipment hatch, personnel airlocks, and penetrations, and subsequently propagates to free-field regions as pressure increases. In addition, the relationship between predicted leakage rates and liner plate failure was investigated. Lower liner strain thresholds result in earlier onset and greater magnitude of leakage, emphasizing the critical role of the liner plate in containment integrity. These findings enhance the understanding of leakage mechanisms and provide a robust framework for more accurate integrity assessments of containment buildings. Furthermore, FE-based leakage prediction methods show strong potential for integration into severe accident codes, enabling a more realistic representation of the relationship between containment leakage rate and internal pressure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114792"},"PeriodicalIF":2.1,"publicationDate":"2026-01-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078930","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 and structural optimization of milliwatt-level radioisotope thermoelectric generators with end-mounted thermoelectric modules","authors":"Hang Jing ,&nbsp;Jing Li ,&nbsp;Xiaoxi Chen ,&nbsp;Qingpei Xiang ,&nbsp;Rende Ze ,&nbsp;Heng Yan ,&nbsp;Liqun Shi ,&nbsp;Shuming Peng","doi":"10.1016/j.nucengdes.2026.114791","DOIUrl":"10.1016/j.nucengdes.2026.114791","url":null,"abstract":"<div><div>Performance and structural optimization of milliwatt-level radioisotope thermoelectric generators (RTGs) with end-mounted thermoelectric modules (TEMs) are investigated. A one-dimensional heat transfer model was developed to analyze temperature distribution and maximum output power (<em>P</em>_<em>max</em>) of the RTG. The sensitivity of <em>P</em>_<em>max</em> to TEM length <em>(L)</em> and cross-sectional area (<em>A</em>) was evaluated for RTGs using five thermoelectric materials. Results show that longer <em>L</em> and optimized <em>A</em> enhance the temperature difference (<em>ΔT</em>) and <em>P</em>_<em>max</em>. For a Bi₂Te₃-based RTG with single end TEM (RTG-1), optimal <em>P</em>_<em>max</em> reached 160.19 mW on Earth at <em>L</em> = 28 mm and <em>A</em> = 292.41 mm², and 273.17 mW on Titan at <em>L</em> = 28 mm and <em>A</em> = 161.29 mm². Dual-end TEM configurations (RTG-2) yielded identical power outputs. COMSOL simulations validated the model with &gt;90% accuracy. Thermal contact resistance (<em>R</em>_<em>C</em>) analysis revealed higher <em>R</em>_<em>C</em> necessitates larger <em>L</em>/<em>A</em> ratios for optimal performance. The model provides a versatile tool for designing RTGs with diverse thermoelectric materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114791"},"PeriodicalIF":2.1,"publicationDate":"2026-01-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078931","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}