Nuclear Engineering and Design最新文献

{"title":"Pre − test evaluation of a grid − spaced deformed fuel pin bundle","authors":"S. Dalvi ,&nbsp;F. Roelofs ,&nbsp;I. Di Piazza ,&nbsp;P.C. Puviani ,&nbsp;V. Moreau ,&nbsp;M. Profir ,&nbsp;M. Fiore ,&nbsp;S. Lopes","doi":"10.1016/j.nucengdes.2025.114293","DOIUrl":"10.1016/j.nucengdes.2025.114293","url":null,"abstract":"<div><div>A code-to-code comparative study is conducted among four contributors (CRS4, ENEA, NRG, VKI) to simulate the Deformed Fuel Pin Simulator of NACIE-UP facility. The objective of the discussed numerical work is to support the planned experiments of the Deformed Fuel Pin Simulator that replicates operating conditions of the ALFRED reactor. For this purpose, a comparative study is carried out using the Computational Fluid Dynamics based framework. Specifically, four distinct numerical approaches are implemented by individual contributors to simulate the operation of the considered test section. In this, blind numerical computations are performed for three different reference test cases, which are later planned for experimental investigation. The influence of pin deformation and of different input parameters, such as the turbulence model and the turbulent Prandtl number, is discussed through independent sensitivity studies. In addition, the effect of distinct mass flow rates and power inputs is observed with the help of several local and global flow variables. The obtained flow-thermal characteristics are then further discussed and compared among all four participants. It is observed that, despite using dissimilar numerical frameworks, thermal measurements are accurately predicted for all considered test cases of low Prandtl number flow. These findings can be helpful to the heavy liquid modelling community that often relies on different numerical tools due to several computational constraints. Insights from the present work are helpful to understand the flow behaviour of liquid metal fast reactor and are intended to be used in a subsequent code-to-experiment comparison.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114293"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144632716","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 heat transfer enhancement mechanism and performance evaluation of streamline spirally coiled tubes","authors":"Chuang Pan ,&nbsp;Xiaoao Li ,&nbsp;Shuhong Li ,&nbsp;Yanjun Li ,&nbsp;Jun Wu ,&nbsp;Gui Li","doi":"10.1016/j.nucengdes.2025.114321","DOIUrl":"10.1016/j.nucengdes.2025.114321","url":null,"abstract":"<div><div>Owing to its advantages of compact structure, large heat transfer area, and high heat transfer efficiency, the coiled tube has been widely used in reactor core cooling and other fields. In this paper, a novel high-efficiency coiled tube is proposed. And the effects of structural parameters on the heat transfer performance and pressure drop of streamline spirally coiled tubes (SSCT) are investigated through numerical simulations. Furthermore, the enhanced heat transfer mechanism is analyzed based on the field synergy theory. The practical value of SSCT is evaluated using the performance evaluation criterion (<em>PEC</em>). The results indicate that as inclination angle (<em>α</em>₁) increases, both the Nusselt number (<em>Nu</em>) and friction factor (<em>f</em>) initially decrease and then increase. As apex angle (<em>α</em>₂) increases, both <em>Nu</em> and <em>f</em> gradually decrease. As twist angle (<em>α</em>) increases, both <em>Nu</em> and <em>f</em> gradually increase. The <em>Nu</em> and <em>f</em> of SSCT (<em>α</em> = 10800°, <em>α</em>₂ = 30°) reach their maximum values, which are 1.44–1.65 times and 2.49–2.61 times those of the circular coiled tube (CCT), respectively. Correspondingly, its <em>PEC</em> is 1.06–1.20, which is consistent with that of SSCT (<em>α</em> = 7200°, <em>α</em>₂ = 30°). This indicates that when <em>α</em> 7200°, SSCT has already achieved a relatively high comprehensive performance. And there is no need to continue increasing the degree of twist. Finally, correlations for the streamline coiled tube (SCT) and SSCT are proposed through linear fitting. The error of the correlation for SCT is within 5 %, while the error for SSCT is within 15 %. This has guiding significance for the practical engineering application of SCT.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114321"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144623696","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":"Scaling methodology and transient distortion evaluation for loop-type sodium-cooled fast reactors","authors":"Lingyi Zhang ,&nbsp;Wenxi Tian ,&nbsp;Peide Zhou ,&nbsp;Yong Yang ,&nbsp;Yufeng Lv ,&nbsp;Suizheng Qiu ,&nbsp;G.H. Su","doi":"10.1016/j.nucengdes.2025.114305","DOIUrl":"10.1016/j.nucengdes.2025.114305","url":null,"abstract":"<div><div>For loop-type sodium-cooled fast reactors employing the sodium-sodium-air configuration, the primary natural circulation path constitutes a single closed loop. While this design facilitates scaled integral tests for validating natural circulation capability, existing methodologies lack dedicated research systematically incorporating the unique characteristics of loop-type SFRs. The scaling criteria were derived through dynamical system scaling modified by the Favre-averaging method. Incorporating the characteristics of loop-type SFR components, the parameter determination methodology was established and subsequently validated using the FFTF reactor as the representative case. The prototypic numerical model was established and validated against the Benchmark for FFTF LOFWOS Test #13, subsequently serving as the foundation for model scaling. Numerical comparisons between the scaled and prototype models were performed for three operational regimes: initial forced circulation steady-state, stable natural circulation, and transient response during LOFWOS Test #13. The analysis investigated three frictional resistance correction approaches and two intermediate heat exchanger parameter design methods, evaluating the impacts on transient process. The transient distortions among different schemes were quantified by the time-value-difference dynamic time warping distortion evaluation code. This study confirms the theoretical feasibility of performing scaled sodium tests for loop-type SFRs with the proposed scaling and distortion evaluation methodology, while also revealing additional distortion factors requiring consideration in practical scaled test implementation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114305"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144623730","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":"A systematic review on grout in nuclear waste management: Advancement, composition and performance","authors":"Anzaman Hossen,&nbsp;Md Munim Rayhan,&nbsp;Md Sharif Ahmed Sarker,&nbsp;Anirban Saha,&nbsp;Abderrachid Hamrani,&nbsp;Dwayne McDaniel","doi":"10.1016/j.nucengdes.2025.114281","DOIUrl":"10.1016/j.nucengdes.2025.114281","url":null,"abstract":"<div><div>Grouts have emerged as a critical solution for the containment of radioactive waste, offering long-term stability, low permeability, and radionuclide immobilization. The safe containment of nuclear waste remains a paramount challenge due to the hazardous and long-lived nature of radioactive materials generated from nuclear power production, medical isotope creation, and military applications. Over the past decades, advancements in grout formulations have led to the development of innovative materials and techniques to enhance their performance in nuclear waste management. This systematic review examines recent progress in grout formulation and performance evaluation by synthesizing data from academic literature. Key focus areas include mechanical properties, permeability reduction, and leaching resistance—parameters essential for ensuring structural integrity and long-term containment. The review highlights the effects of incorporating materials such as silica fume, fly ash, and slag to improve durability and chemical stability. Additionally, innovations like flowable, zero-bleed grouts are explored for their potential in withstanding extreme environmental conditions. Through performance assessments and material characterization, this review identifies prevailing research trends and formulation strategies, offering insights into the optimization of grout systems for nuclear waste management applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114281"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144623731","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 and verification of thin-walled particle buffer under transient impact load","authors":"Wangqiang Xiao ,&nbsp;Qibin Liu ,&nbsp;Zhiqin Cai ,&nbsp;Zhaoyang Liu ,&nbsp;Guilong Wang ,&nbsp;Yu Dai ,&nbsp;Shengbo Wang","doi":"10.1016/j.nucengdes.2025.114304","DOIUrl":"10.1016/j.nucengdes.2025.114304","url":null,"abstract":"<div><div>As a key safety component of the spent fuel transport cask, the cushioning performance of the buffer is essential for ensuring safe transportation of spent fuel, and this performance depends on the buffer material used. Currently, continuous materials such as wood provide buffering effects solely through elastic–plastic deformation energy dissipation. To overcome this limitation, this paper proposes a design method for thin-walled particle buffers in spent fuel transport casks. Based on the finite element-discrete element coupling theory, this paper establishes a continuous-discontinuous coupling model for the buffer of the spent fuel transport cask. The buffer shell of the continuous medium is simulated using the finite element method (FEM), while the particle system of the discontinuous medium is modeled via the discrete element method (DEM). This approach enables equivalent mapping of contact loads from the DEM domain to the FEM domain at the nodal level. This method has addressed the limitation that a single discrete element method (DEM) or finite element method (FEM) cannot fully solve the system’s dynamic response. By filling the buffer cavity with thin-walled particle media, we constructed a comprehensive energy dissipation superposition mechanism combining damping dissipation and elastoplastic energy attenuation. The influence of the characteristic parameters of the buffer on the buffering performance was explored, and the parameter scheme of the thin-walled particle medium with the best buffering performance was obtained. A drop impact test platform was established to conduct drop tests on all schemes and compare them with the simulation calculation results, verifying the accuracy of the coupling model. Finally, the 9-meter drop test of the optimal buffering performance scheme was verified. The similarity theory calculated that the prototype’s acceleration under this working condition was 56.1 g, the bolt shear stress was 215 MPa, and the axial stress was 867 MPa. The containment boundaries all met the relevant standard requirements. This paper’s research can provide new ideas for the design of buffers in the spent fuel transport casks.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114304"},"PeriodicalIF":1.9,"publicationDate":"2025-07-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144604770","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 thermal characteristics and enhancement measures for the heat exchange system in the pressurized water test loop during fuel assembly irradiation under off-design conditions","authors":"Junping Si,&nbsp;Guang Zhao,&nbsp;Yun Wang,&nbsp;Mingyan Tong,&nbsp;Sheng Sun,&nbsp;Hongwei Wu,&nbsp;Yong Luo,&nbsp;Wenlong Zhang,&nbsp;Shuai Jin,&nbsp;Jiaxin He","doi":"10.1016/j.nucengdes.2025.114315","DOIUrl":"10.1016/j.nucengdes.2025.114315","url":null,"abstract":"<div><div>The thermal characteristics of heat exchange systems under off-design conditions for fuel assembly irradiation tests in a pressurized loop were analyzed, and heat transfer enhancement methods to address insufficient heat exchange capacity were further compared. It is shown that the power variation of the primary heat exchanger is closely related to the variation in primary water flow rate and shows a positive linear relationship. Meanwhile, due to the limitations of secondary water conditions, there is a risk of insufficient heat exchange capacity for heat exchangers designed with a fixed structure and parameters under off-design conditions. Besides, when the heat exchanger operates at low power under off-design conditions, adjusting the secondary water flow rate has little effect on improving the main heat exchanger’s performance. Using lower-temperature secondary water and increasing the primary water inlet temperature can partially improve the heat exchanger’s power. The rear-end cooling method has a higher total heat exchange capacity than front-end cooling. Moreover, the heat exchange effect of parallel heat exchangers is not always superior to that of a single unit. Under low-flow conditions, the most effective solution to insufficient heat exchange power is to use multiple heat exchangers with the ability to operate in series. For primary water inlet temperatures of 330 °C and 250 °C, the total heat exchange power of two heat exchangers in series increases by 76.6 % and 80.4 % compared to single-unit operation, respectively.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114315"},"PeriodicalIF":1.9,"publicationDate":"2025-07-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144604265","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":"Ultrasonic Doppler velocimetry (UDV) enabled velocity field measurements in a liquid metal plenum with colder jet injection","authors":"Broderick Sieh,&nbsp;Hitesh Bindra","doi":"10.1016/j.nucengdes.2025.114273","DOIUrl":"10.1016/j.nucengdes.2025.114273","url":null,"abstract":"<div><div>Liquid-metal cooled nuclear reactor systems experience thermal-fluid transients which can severely impact safety analyses. These transients can pose multiphysics challenges which need to be carefully studied to further enhance the understanding of reactor safety. The most significant obstacle impeding the improvement of this understanding is the complicated task of obtaining high-fidelity temperature and flow field measurements in complex liquid-metal reactor systems.</div><div>The Gallium Thermal-hydraulic Experiment (GaTE) is an experimental setup aimed at investigating transient conditions in liquid metal (LM) cooled reactors using a scaled-down model of the reactor outlet plenum. Gallium is used in place of sodium as it also exhibits a low Prandtl number (<span><math><mrow><mi>P</mi><mi>r</mi></mrow></math></span>) in its liquid state. Moreover, gallium is affordable, non-toxic, and non-reactive with both air and water, unlike sodium. To replicate the required transient conditions in the scaled-down plenum, GaTE can control the velocity of the inlet jet, the pool temperature, and the overall system circulation. Thermocouples are used to monitor the temperature of the gallium throughout the loop. As the plenum is the component in such reactors expected to experience the most complex thermal-fluid behavior, a fiber optic distributed temperature sensing (DTS) system aligned in a multi-pass array is installed inside the plenum to measure the temperature of the gallium pool.</div><div>In prior studies, GaTE has been employed to investigate thermal stratification in a scaled LM cooled reactor plenum, focusing on the Richardson number (<span><math><mrow><mi>R</mi><mi>i</mi></mrow></math></span>). The tests made use of the DTS system to characterize thermal stratification and the oscillations linked with the stratification thermocline. Connecting the flux parameters to global parameters necessitates gathering flow velocity data inside the plenum along with temperature data. Ultrasonic Doppler velocimetry (UDV) is applied to acquire the necessary data during cold shock transients within the scaled LM cooled reactor plenum. Through employing UDV, the effects of Reynolds number (<span><math><mrow><mi>R</mi><mi>e</mi></mrow></math></span>) and temperature on LM flow profiles within the plenum are assessed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114273"},"PeriodicalIF":1.9,"publicationDate":"2025-07-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144588113","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":"Ar-41 source term investigation in HTR-PM: Theory, calculations, and experiments","authors":"Rui Nie ,&nbsp;Jiong Guo ,&nbsp;Qian Ma ,&nbsp;Xi Chen ,&nbsp;Yulin Zhang ,&nbsp;Wenting Jia ,&nbsp;Yu Wang ,&nbsp;Feng Xie ,&nbsp;Jianzhu Cao ,&nbsp;Yaoquan Zhou ,&nbsp;Emmanuel K. Boafo ,&nbsp;Qingming He ,&nbsp;Liangzhi Cao","doi":"10.1016/j.nucengdes.2025.114271","DOIUrl":"10.1016/j.nucengdes.2025.114271","url":null,"abstract":"<div><div>Ar-41 is a significant activation product in high-temperature gas-cooled reactors (HTGRs). Based on the design and operational parameters of a high-temperature gas-cooled reactor pebble-bed module (HTR-PM), we analyzed the Ar-41 source and constructed an Ar-41 source-term calculation model for HTGRs. For the first time, we presented the measured data of Ar-41 activity in the primary coolant of HTR-PM: during startup, the experimental value could reach up to 10⁹ Bq·m⁻³, consistent with the theoretical prediction of 1.99×10⁹ Bq·m⁻³; during normal operation, that fluctuated between 10⁶ and 10⁸ Bq·m⁻³, in agreement with the theoretical value of 3.27×10⁷ Bq·m⁻³. The effective dose rates for workers in the reactor, steam generator, helium purification system, and fuel loading and unloading compartments are 4.87×10^-2, 3.17×10^-4, 2.17×10^-5, and 9.23×10^-6 Sv·d^-1, respectively. The largest effective dose rate for adult members of the public within an 80 km radius from the plant is 6.61×10^-8 Sv·d^-1. In addition, the key factors that can affect the Ar-41 activity in the primary coolant and the effect of Ar-41 on the activity concentration conversion factor of the gross γ monitoring instrument were discussed. This work provides an important reference for studying the radiation safety in advanced nuclear energy systems and fusion reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114271"},"PeriodicalIF":1.9,"publicationDate":"2025-07-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144587430","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":"Flow and heat transfer characteristics of He-Xe gas mixture and H2O in printed circuit heat exchanger (PCHE)","authors":"Haoran Shen ,&nbsp;Yang Ming ,&nbsp;Ruibo Lu ,&nbsp;Ersheng You ,&nbsp;Fulong Zhao ,&nbsp;Sichao Tan","doi":"10.1016/j.nucengdes.2025.114298","DOIUrl":"10.1016/j.nucengdes.2025.114298","url":null,"abstract":"<div><div>The He-Xe Brayton cycle system is currently one of the most feasible technical solutions for small modular reactor power systems, where the thermo-hydraulic performance of its cooler plays a critical role in determining the energy conversion efficiency. Due to the significant differences in physical properties between the working fluids on both sides of the cooler, traditional cooler structures face limitations, necessitating structural optimization. This study proposes a novel cooler configuration and investigates the flow and heat transfer characteristics of He-Xe mixture and H₂O in straight-channel printed circuit heat exchanger (PCHE) using numerical simulation method. The results indicate that both temperature and velocity fields within the channels follow exponential distributions. The average convective heat transfer coefficient of H₂O is approximately three times that of He-Xe, while the average Nusselt number of He-Xe exceeds H₂O by 109 %. Furthermore, variations in the hot-side (He-Xe) inlet temperature and mass flow rate exert greater influence on PCHE heat transfer performance compared to cold-side (H₂O) inlet temperature adjustments. Specifically, every 50 K increase in He-Xe inlet temperature enhances heat transfer by 96 %, and a 0.5 g/s increment in mass flow rate boosts heat exchange power by 25 %. Increasing the operating pressure of He-Xe effectively reduces pressure drop while maintaining heat exchanger performance. This study can provide reference for the design and operation of PCHE heat exchangers in He-Xe Brayton cycle reactor systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114298"},"PeriodicalIF":1.9,"publicationDate":"2025-07-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144587429","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":"Modeling of the multi-scale related thermo-mechanical coupling behaviors in a helical cruciform fuel assembly","authors":"Xiaoxiao Mao ,&nbsp;Xingdi Chen ,&nbsp;Xiaobin Jian ,&nbsp;Feng Yan ,&nbsp;Tenglong Cong ,&nbsp;Yao Xiao ,&nbsp;Hui Guo ,&nbsp;Hanyang Gu ,&nbsp;Shurong Ding","doi":"10.1016/j.nucengdes.2025.114276","DOIUrl":"10.1016/j.nucengdes.2025.114276","url":null,"abstract":"<div><div>The U-Zr based helical cruciform fuel (HCF) assemblies exhibit great potential for application in advanced future reactors, due to their high thermal conductivity, high uranium density, large heat transfer surface area, and “self-spacing” characteristics. In this study, a mesoscale normal stress model for porous fuel is developed based on a novel volumetric growth strain model of U-70at%Zr alloy. In this study, a mesoscale normal stress model for porous fuel is developed based on the novel volumetric growth strain model of U-70at%Zr alloy. Considering the unique geometry of HCF rods and the anisotropic irradiation growth strains of Zr-4 cladding in the material principal ordinate system, an anisotropic irradiation growth strain increments has been established in the defined co-rotational coordinate system. Hence, numerical simulation of the thermo-mechanical coupling behaviors in U-70at%Zr-based HCF rods and assemblies are conducted. The numerical simulation results indicate that: (1) a 1/3 simplified fuel assembly model with a 1/4 twist pitch length effectively captures the overall thermo-mechanical behaviors of the entire fuel assembly; (2) At a fission density of 1.73 × 10²¹ fissions/cm³, the maximum temperature at the fuel rod core center reaches approximately 632.5 K; the maximum lateral displacement of the fuel assembly is about 1.35 mm, with minimal impact on the coolant channels; thermal dislocation creep and irradiation creep deformations in the solid skeleton of the porous fuel rod core significantly influence gas bubble growth; (3) over a 2000-day irradiation period, the maximum contact pressure and the maximum absolute value of tangential contact stress are 98.1 MPa and 8.9 MPa, respectively; (4) the mesoscale normal stress for fuel skeleton is only 1.2 MPa on the 400th day, increasing to 39.8 MPa and 38.2 MPa on the 1600th day and 2000th day; the maximum von Mises stress is ∼ 189 MPa at the end structures, with no equivalent plastic strain produced. This research can provide a reference for the advanced manufacture and the optical design of HCF assembly with U-Zr fuel rod core and Zr-4 cladding.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114276"},"PeriodicalIF":1.9,"publicationDate":"2025-07-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144588105","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}