Nuclear Engineering and Design最新文献

{"title":"Developed high-strength high-ductility 46.7 GPa.% austenitic stainless steel as fuel cladding in fast nuclear reactor","authors":"Sara E. Saleh , M.K. Elfawakhry , R.M. El Shazly , Heba A. Saudi , S.M. El-Minyawi , M.M. Eissa","doi":"10.1016/j.nucengdes.2025.114080","DOIUrl":"10.1016/j.nucengdes.2025.114080","url":null,"abstract":"<div><div>Four samples of austenitic stainless steel were prepared to study the effect of adding titanium, and nickel/chromium modification on the characteristic properties of ordinary austenitic stainless steels, AISI304L and AISI316L that are used in a fuel cladding in fast breeder reactor and the characteristic properties of the developed steels have been compared with the standard alloys. Thermo-calc program FEDAT database was used to predict the phases that can be formed in the different alloys from room temperature to elevated temperature. The constituent phases have been detected by scanning electron microscope attached with EDS and X-ray diffraction. The mechanical properties of investigated stainless-steel alloys were monitored through using uniaxial tensile test, and impact resistance. Corrosion resistance of the studied stainless-steel alloys were investigated in 3.5 % NaCl solution to determine their corrosion rate. The results refer to that the modified austenitic stainless-steel samples with nickel increment at the expense of chromium and micro-alloyed with titanium have preferable mechanical properties in comparison with the standard austenitic stainless-steels AISI316L and AISI304L The yield strength of the developed stainless-steel alloys is enhanced by 21 % and 4 % compared to the standard SS304L and SS316L alloys, respectively. This directly improves the material’s ability to endure extreme conditions, ensuring greater reactor safety, longevity, and performance. The developed SS304LTi showed the best combined high-strength and high-ductility with 46.7 GPa.%. In addition, Furthermore, the corrosion rates of the developed stainless-steel alloys were found to be 58 % and 41 % lower than those of the standard SS304L and SS316L alloys, respectively. This reduction is highly significant, particularly in terms of safety, durability, and the overall efficiency of the reactor.</div><div>To investigate the accommodate of the developed stainless steels in structure of nuclear reactor, four different types of neutron energies were used to determine the macroscopic neutron cross-sections (Σ, cm-1) for the prepared stainless-steel alloys and mean free path was calculated. WinX-com computer program (Version 3.1), and nine experiments of different gamma ray energy lines up to 1.4 MeV were used to determine the mass attenuation coefficients (σ, cm2/g) of gamma rays for the prepared stainless-steel alloys. Good agreement was found between the experimental and calculated values of mass attenuation coefficient. The developed SS304LTi and SS316LTi austenitic stainless steels have lower HVL comparing with the standard SS304L and SS316L, and consequently higher effectiveness of shielding material at the related photon energy. Furthermore, the developed SS304LTi and SS316LTi austenitic stainless steels showed greater values of macroscopic cross-sections and lower values of MFP in all types of neutron energies comparing with the standard SS304L and SS3","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114080"},"PeriodicalIF":1.9,"publicationDate":"2025-04-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143843082","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 flow regime transitions and frictional pressure drop in a 9x9 helical cruciform fuel bundle","authors":"Matthew Kinsky ,&nbsp;Hansol Kim ,&nbsp;Dalton W. Pyle ,&nbsp;Joseph Seo ,&nbsp;Yassin A. Hassan","doi":"10.1016/j.nucengdes.2025.114074","DOIUrl":"10.1016/j.nucengdes.2025.114074","url":null,"abstract":"<div><div>This study experimentally investigates the frictional pressure loss and flow regime behavior of a 9 × 9 Helical Cruciform Fuel (HCF) rod bundle, a novel design proposed for Small Modular Reactors (SMRs). The unique cruciform cross-section, featuring four twisted petals, eliminates the need for conventional spacer grids, offering higher fuel packing fraction and enhanced coolant mixing. To assess these advantages, a high-precision differential pressure measurement system was employed over a Reynolds number range of 200–22,000, covering the laminar, transition, and turbulent flow regimes. The experimentally determined friction factors showed statistically similar trends between the “one pitch” and “bundle-averaged” axial segments, confirming fully developed flow in both regions. Empirical correlations for friction factor and differential pressure per unit length were then developed for each flow regime and validated by comparison with previous HCF and wire-wrapped fuel bundle studies. Results identified flow regime boundaries at approximately Re ≈ 1000 for laminar-to-transition and Re ≈ 8274 for transition-to-turbulent, highlighting distinctly different hydraulic behavior in the three regimes. The findings significantly broaden the limited experimental database on HCF rod bundles, providing new insights into regime-dependent pressure drop characteristics. By refining existing correlations and offering high-fidelity benchmark data, this work advances the development of more efficient and accurate reactor core designs that leverage HCF technology for enhanced thermal performance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114074"},"PeriodicalIF":1.9,"publicationDate":"2025-04-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143838888","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":"Effect of phase transformation on the high-temperature tensile behaviors of SA508 Gr. 3 steel: A crystal plasticity finite element investigation","authors":"Silu Zheng ,&nbsp;Haolin Yu ,&nbsp;Xiatao Tang ,&nbsp;Jiahe Zhou ,&nbsp;Chuanyang Lu ,&nbsp;Yuebing Li ,&nbsp;Yanming He ,&nbsp;Zengliang Gao","doi":"10.1016/j.nucengdes.2025.114070","DOIUrl":"10.1016/j.nucengdes.2025.114070","url":null,"abstract":"<div><div>The in-vessel retention (IVR) strategy, designed to maintain the structural integrity of reactor pressure vessels (RPVs) during severe nuclear accidents, will induce a huge temperature gradient across the RPV wall. This temperature gradient may lead to an austenitic phase transformation within RPV materials. Due to the dual-phase microstructure caused by this phase transformation, predicting high-temperature mechanical properties, e.g. tensile strength, becomes challenging, thereby impeding the implementation of IVR for RPVs. In this work, crystal plasticity finite element method (CPFEM) coupled with austenite transformation kinetics (ATK) was employed to model the tensile behaviors of SA508 Gr.3 steel, a typical RPV material, at three stages: 1) before phase transformation with ferrite phase (700–973 K), 2) during phase transformation with dual phases (973–1073 K) and 3) after phase transformation with austenite phase (1073–1273 K). The results demonstrate that stress concentrations primarily occur at a deflection of 140–150° between the normal direction of slip plane and loading direction in both ferrite and austenite grains, consistent with Schmid’s law. In materials undergoing phase transformation, the locations of stress-concentrated grains and their stress distributions are influenced by: 1) deflection angle, 2) grain type, and 3) misorientation angles between neighboring grains. The tensile behaviors during phase transformation with dual phases are predicted using this CPFEM-ATK method. These findings will provide comprehensive insights into the high-temperature tensile behaviors of RPV materials in IVR conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114070"},"PeriodicalIF":1.9,"publicationDate":"2025-04-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143834192","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":"Insights on the instability and stabilizing techniques for natural circulation loops","authors":"P.K. Vijayan ,&nbsp;Swati Gangwar ,&nbsp;Dev Banitia ,&nbsp;U.C. Arunachala ,&nbsp;S. Nakul ,&nbsp;D.N. Elton ,&nbsp;K. Varun","doi":"10.1016/j.nucengdes.2025.114017","DOIUrl":"10.1016/j.nucengdes.2025.114017","url":null,"abstract":"<div><div>There is a generally held belief that the insertion of an orifice which is equivalent to increasing the L_t/D ratio is always stabilizing SPNCSs. In this paper, it has been shown that the insertion of an orifice can stabilize or destabilize depending on whether the loop is operating near the lower or upper threshold of instability for single-phase loops. Besides, increasing the L_t/D ratio increases the unstable zone in single-phase loops and, hence, is destabilizing. For two-phase loops, insertion of an orifice or increasing the L_t/D ratio significantly shrinks the stable zone increasing the unstable zone as in single-phase loops. Thus for both single-phase and two-phase loops, reducing the L_t/D is stabilizing. Contrary to this, for the supercritical loops L_t/D ratio (or orificing) has a complex effect on instability. For example, increasing the L_t/D or insertion of an orifice shrinks the unstable zone giving a stabilizing effect. Also, reducing the L_t/D ratio is seen to shift both the lower and upper thresholds to higher powers and, in this sense, is stabilizing. However, it is also found to widen the unstable zone with a decrease in L_t/D and, in this sense, is destabilizing.</div><div>The paper also reviews the available stabilizing techniques to identify the techniques which do not significantly reduce the heat transport capability while stabilizing. For single-phase and two-phase loops, the best way to stabilize is the reduction of L_t/D ratio as it stabilizes with enhancement in heat transport capability. Introduction of an orifice enhances the unstable zone in single-phase and two-phase loops whereas it has a mixed effect in supercritical loops. Increase in L_t/D is found to reduce the flow and hence narrows down the pseudocritical region and hence the unstable region to stabilize supercritical loops. Reduction of L_t/D ratio is found to stabilize supercritical loops at high inlet temperatures, whereas it widens the unstable region at low inlet temperatures, which is attributed to the widening of the pseudocritical region. The paper also examines the various requirements for maximizing the power of natural circulation based reactors. Apart from reducing the frictional force, enhancing the surface area density in the core has a significant influence on enhancing the reactor power and various options for the same has been identified in the paper.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114017"},"PeriodicalIF":1.9,"publicationDate":"2025-04-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143834194","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":"Neutronic design of a novel small modular reactor based on the dual-cooled accident tolerant fuels using systematic methodology: Fuel assembly and core pattern evaluation via artificial neural network","authors":"H. Zayermohammadi Rishehri ,&nbsp;G.R. Ansarifar ,&nbsp;M. Zaidabadi Nejad","doi":"10.1016/j.nucengdes.2025.114057","DOIUrl":"10.1016/j.nucengdes.2025.114057","url":null,"abstract":"<div><div>This study investigates the design of a novel Small Modular Reactor (SMR) concept utilizing Dual-Cooled Accident Tolerant Fuel (DC-ATF). The DC-ATF incorporates U₃Si₂ fuel pellets clad in FeCrAl, enhancing safety and accident tolerance. A systematic approach was employed, beginning with the evaluation of 4000 unique fuel assembly configurations varying the number and arrangement of Integrated Burnable Absorbers (IBAs). Fifty configurations in each category were rigorously simulated using the MCNP code, and the results were used to train Artificial Neural Networks (ANNs) to predict the performance of the remaining assemblies. This approach facilitated the identification of suitable fuel assembly designs for each IBA category. Subsequently, these assemblies were integrated into 55 distinct reactor core configurations, varying the distribution of IBA-containing assemblies within a 37-assembly core arranged in a square lattice. Neutronic simulations were performed to evaluate core criticality, power distribution, burnup characteristics, and temperature coefficients. The results demonstrate that the proposed DC-ATF SMR exhibits favorable safety margins, including negative temperature coefficients (−2 pcm/K for fuel and −33.89 pcm/K for coolant) and acceptable power peaking factors (1.58 at beginning of the cycle). Burnup calculations indicate a first core cycle length exceeding 1800 effective full power days (EFPD), a significant increase compared to conventional UO₂-fueled SMRs of similar size and output power, which typically achieve around 730–1330 EFPD. This improvement is primarily attributed to the higher uranium density of U₃Si₂ fuel, enabling increased fissile material loading.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114057"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143830245","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":"New correlation for Critical heat flux in annuli with low pressure water at low flow rates","authors":"Mirza M. Shah","doi":"10.1016/j.nucengdes.2025.114072","DOIUrl":"10.1016/j.nucengdes.2025.114072","url":null,"abstract":"<div><div>Calculation of CHF (Critical Heat Flux) for low pressure low flow water in annuli is required in the design and analyses of conventional nuclear reactors as well as the newer advanced nuclear reactors. Correlations for high pressure high flow have been found to fail for low pressure low flow conditions. There are no well-verified correlations under these conditions. The few published correlations have been verified with only a limited amount of data. In the present research, these correlations were compared to all available data which was from many sources. None of them was found satisfactory. A new correlation was therefore developed which agrees well with all available data for vertical annuli with upflow. The range of data included annular gaps 0.9 to 16.5 mm, pressures 1 to 3.2 bar, mass flux 1 to 1030 kg/m²s, and inlet quality −0.17 to 0. The new correlation had MAD (mean absolute deviation) of 19.6 % with 273 data points from 13 sources. The MAD of other correlations ranged from 45.3 % to 99.7 %. In this paper, previous work is reviewed, development of the new correlation is described, and comparison of the new and earlier correlations with test data is presented. Recommendations are made for its application.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114072"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143826161","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":"Inclined projectile impact on reinforced concrete structures","authors":"Lars Heibges,&nbsp;Hamid Sadegh-Azar","doi":"10.1016/j.nucengdes.2025.114043","DOIUrl":"10.1016/j.nucengdes.2025.114043","url":null,"abstract":"<div><div>Impact loads, such as airplane or debris crashes, are a significant load case in the safety assessment and design of nuclear facilities. In the past, research on impact events primarily focused on impact scenarios with normal angle. However, in real-world situations, an inclined angle of impact can be expected for impact events. As a result, there is a growing need to investigate the effects of inclined impact on reinforced concrete structures with a focus on the resulting damage and failure modes. Understanding the effects of impact angles on the load-bearing capacity of these structures is crucial for ensuring their safety and integrity.</div><div>This paper examines the effects of inclined projectile impacts on the load-bearing capacity of reinforced concrete structures for both soft and hard missiles. Nonlinear dynamic numerical simulations using 3D fully coupled analysis are conducted and validated against experimental test results from the literature. Different friction models are implemented and evaluated for punching and bending responses. The friction models examined in this paper show strong agreement with experimental data, confirming their reliability in simulating both punching and bending tests.</div><div>In addition to the numerical analyses, simplified approaches for calculating the support forces as well as residual velocities for different impact angles are investigated and validated with experimental data and simulations, showing reasonable agreement with both numerical models and experimental data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143826045","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":"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}