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":"Experimental investigation of gas-phase migration behaviors in a 1 × 2 rod bundle under two-phase equilibrium and non-equilibrium flow","authors":"Hao Xie,&nbsp;Wenhai Qu,&nbsp;Jinbiao Xiong","doi":"10.1016/j.nucengdes.2026.114806","DOIUrl":"10.1016/j.nucengdes.2026.114806","url":null,"abstract":"<div><div>Void fraction gradient was assumed as the mechanism of void drift between subchannels. However, classical void drift models based on this assumption predict poor results under bubbly flow and cap-bubbly flow against experimental data. In this work, void fraction and bubble diameter in an enlarged 1 × 2 rod bundle was experimentally investigated by wire mesh sensors (WMS) system under equilibrium and non-equilibrium gas flow rates. The total 168 cases cover bubbly flow, cap-bubbly flow and churn flow defined with bubble shape and volume-base probability density function (VPDF). In general, small bubbles gather near channel walls, while large bubbles concentrate in sub-channel centers. For bubbly flow and cap-bubbly flow, it is difficult for bubbles passing through gap between subchannels. Thus, classical void drift models deviate from experimental results in bubbly flow and cap-bubbly flow. However, when bubble diameter is larger than 0.75 time of pitch of rod bundle, void drift improves obviously because bubbles are large enough to disturb the corresponding subchannel. Under bubbly flow and cap-bubbly flow, large bubbles coalesce in each subchannel. With gas superficial velocity increasing, void fraction and bubble diameter of large bubbles increase, while VPDF of small bubbles decreases. With liquid superficial velocity increasing, void fraction and bubble diameter of large bubbles decrease, while VPDF of small bubbles increases. For churn flow, strong void fraction migration between sub-channels is mainly caused by large bubbles with diameter larger than pitch of 32 mm. Thus, equilibrium distribution of void fraction can be fully developed within distance of 1830 mm under equilibrium and non-equilibrium inlet conditions. This is the reason that classical void drift models predict well against experimental data in churn flow. Large bubbles break down in churn flow. With gas superficial velocity increasing, void fraction and bubble diameter of large bubbles increase, while VPDF of small bubbles decreases. With liquid superficial velocity increasing, void fraction and bubble diameter of large bubbles decrease, while VPDF of small bubbles increases. This work helps to under mechanism of gas migration between sub-channels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114806"},"PeriodicalIF":2.1,"publicationDate":"2026-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190735","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 316L-based neutron shielding materials and preparation and characterization of Gd/316L","authors":"Chunguang Qiao ,&nbsp;Zhonghua Wang ,&nbsp;Xinpeng Wei ,&nbsp;Dehui Wu ,&nbsp;Chao Jiang","doi":"10.1016/j.nucengdes.2026.114808","DOIUrl":"10.1016/j.nucengdes.2026.114808","url":null,"abstract":"<div><div>The corrosion resistance and structural mechanical properties of neutron shielding materials such as B₄C/Al and boron‑aluminum alloy are insufficient. The 316L stainless steel commonly used in the nuclear industry has excellent structural mechanical properties and corrosion resistance, but its neutron shielding performance is poor. The above-mentioned materials are unable to meet the multi-dimensional functional requirements of the core shielding layer of fast reactors, the structural components within the reactor, and the storage of spent fuel, which include high load-bearing capacity, efficient neutron shielding, and excellent corrosion resistance. Therefore, 316L-based neutron shielding materials are designed and prepared using 316L stainless steel as the substrate and B₄C, Gd, Sm₂O₃, and Eu₂O₃ as neutron shielding enhancers. Firstly, the relationships between the areal densities of B₄C, Gd, Sm₂O₃, and Eu₂O₃, and the neutron shielding rate and secondary γ-ray production rate of the corresponding shielding materials are established. Secondly, using the established relational equation, the required contents of B₄C, Gd, Sm₂O₃, and Eu₂O₃ in 316L stainless steel are calculated and compared using the neutron shielding rate of 30% B₄C/Al composite as the design basis. Finally, Gd/316L neutron shielding materials are prepared by directed energy deposition additive manufacturing (DED-AM) process, and their micro-morphology and mechanical properties are analyzed. The results show that Gd is more suitable as a neutron shielding enhancer for 316L stainless steel. The 1.9% Gd/316L exhibits good mechanical properties, while a further increase in Gd content degrades the mechanical performance of the material</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114808"},"PeriodicalIF":2.1,"publicationDate":"2026-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190736","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":"Enhancing small modular reactor adoption for sustainable energy transition: a Fermatean fuzzy ISM-MICMAC framework for analyzing challenges","authors":"Ismail Erol ,&nbsp;Ahmet Oztel ,&nbsp;Ihsan Tolga Medeni ,&nbsp;Ilker Murat Ar","doi":"10.1016/j.nucengdes.2026.114805","DOIUrl":"10.1016/j.nucengdes.2026.114805","url":null,"abstract":"<div><div>Small modular reactors (SMRs) promise reduced upfront costs, faster construction, and enhanced safety compared to traditional reactors. However, widespread adoption is hindered by challenges such as high capital costs, regulatory delays, supply chain inefficiencies, cybersecurity risks, nuclear waste management, and public skepticism. Despite qualitative studies highlighting these barriers, quantitative analyses remain scarce, necessitating systematic frameworks to model interdependencies and guide solutions. The goal of this study is to scrutinize SMR adoption challenges using a novel multi-criteria decision-making (MCDM) approach. Drawing on a literature review from Web of Science, Scopus, and various reports from international institutions, 13 key challenges were identified. A panel of 58 experts—academics, government officials, and cybersecurity specialists—provided inputs via pairwise comparisons. The methodology used in this study integrates Fermatean Fuzzy Interpretive Structural Modeling (FFISM) with the cross-impact matrix multiplication applied to classification (MICMAC) analysis. Fermatean fuzzy sets extend traditional ISM by accommodating higher uncertainty in expert judgments through expanded membership/non-membership degrees. Validation involved 10,000 simulations comparing FFISM to conventional fuzzy ISM. Results reveal a six-level hierarchy: licensing/regulatory constraints and lack of proven technology/FOAK units as top challenges, influencing linkage challenges such as supply chain effectiveness, cybersecurity risks, and waste management. Dependent challenges include perceived investment risk, cost estimation, and public opinion. Policy recommendations include risk-informed licensing to cut timelines, blockchain for traceability addressing fuel availability, and public-private partnerships with green bonds to mitigate risks. This research provides actionable strategies for policymakers and stakeholders to accelerate SMR deployment, strengthening nuclear energy's role in global decarbonization.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114805"},"PeriodicalIF":2.1,"publicationDate":"2026-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190729","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":"Corrigendum to \"Verification of PWR-Core power distribution based on precisely calculated SPND response currents\" <[ Nuclear Engineering and Design 448 (2026) 114726]>","authors":"Sipeng Du,&nbsp;Yunzhao Li,&nbsp;Hangqi Zhang,&nbsp;Ruizhi Shao,&nbsp;Liangzhi Cao","doi":"10.1016/j.nucengdes.2026.114804","DOIUrl":"10.1016/j.nucengdes.2026.114804","url":null,"abstract":"","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114804"},"PeriodicalIF":2.1,"publicationDate":"2026-01-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189362","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}