Nuclear Engineering and Design最新文献

{"title":"Cavitation-assisted jet decontamination of lead–bismuth eutectic alloy from stainless steel surface with using bionic snapping shrimp’s nozzle","authors":"Feng Liu ,&nbsp;Jiawen Li ,&nbsp;Yunkang Bai ,&nbsp;Jinlong Heng ,&nbsp;Xuefei Li ,&nbsp;Cailun Wang ,&nbsp;Jie Zhan ,&nbsp;Chen Zhang","doi":"10.1016/j.nucengdes.2025.114038","DOIUrl":"10.1016/j.nucengdes.2025.114038","url":null,"abstract":"<div><div>The low-melting and high-boiling lead–bismuth eutectic alloy is used in nuclear reactor as heat exchange medium, which causes lead–bismuth eutectic alloy to be adhered on stainless steel surface of nuclear reactor. The current cleaning methods are challenging to remove lead–bismuth eutectic alloy on stainless steel surface with high efficiency and high decontamination effectiveness in an environmentally friendly manner. Thus, a new cavitation-assisted decontamination method with using bionic jet nozzle mapping from snapping shrimp claw is proposed in the paper. Firstly, a new bionic jet nozzle mapping from snapping shrimp claw is designed and manufactured based on the characteristics of cavitation effect generated by the high-speed closure of the claw of snapping shrimp. Secondly, the cavitation-assisted decontamination of lead–bismuth eutectic alloy from stainless steel surface by using the bionic jet nozzle is investigated experimentally and an experimental platform is established with combining the bionic jet nozzle. Finally, the results of decontamination effectiveness are evaluated from comprehensive score with three indicators, observation of surface microscopic morphology, analysis of water quality for waste liquid, solid impurity analysis. 316L stainless steel adhered with lead–bismuth alloy is taken as experimental samples and the decontamination area ratio, weight reduction rate, and surface roughness <em>R_a</em> of the sample are selected as evaluation indicators of decontamination effect. The weights of each indicator are assigned using the entropy weight method and a comprehensive score of the decontamination effect for each sample is calculated by the above method. The optimal process parameters are determined and validated as nozzle inlet pressure 17.5 MPa, target distance 6 mm, and jet duration 8 min. The average values of the decontamination area ratio, weight reduction rate, and surface roughness <em>R_a</em> of the sample after decontamination under the optimal process parameter are respectively 89.9 %, 90.7 %, and 0.816 μm. The test results of water quality for the waste liquid after decontamination show that the chemical oxygen demand of the waste liquid generated by the cavitation-assisted jet decontamination is much lower than the that of chemical immersion decontamination method. This study indicates that the cavitation-assisted jet decontamination method with bionic nozzle mapping from snapping shrimp claw can effectively remove lead–bismuth alloys from the surface of stainless steel and is more environmentally friendly.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114038"},"PeriodicalIF":1.9,"publicationDate":"2025-05-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143934672","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 control strategies for rapid load decrease events in fluoride salt reactor-supercritical CO2 Brayton cycle systems","authors":"Yun Shichang ,&nbsp;Li Xinyu ,&nbsp;Zhang Dalin ,&nbsp;Song Ping ,&nbsp;Tian Wenxi ,&nbsp;Qiu Suizheng ,&nbsp;Su Guanghui","doi":"10.1016/j.nucengdes.2025.114114","DOIUrl":"10.1016/j.nucengdes.2025.114114","url":null,"abstract":"&lt;div&gt;&lt;div&gt;This study delves into the control strategies for the Fluoride-salt-cooled High-temperature Reactor-Supercritical CO&lt;sub&gt;2&lt;/sub&gt; (SCO&lt;sub&gt;2&lt;/sub&gt;) Brayton cycle power generation system under rapid grid load reduction scenarios. By developing a dynamic simulation model, the system’s dynamic response characteristics under rapid load changes were analyzed, and a combined control strategy emphasizing safety, speed, and economic efficiency was proposed.&lt;/div&gt;&lt;div&gt;The research first compared the dynamic response characteristics and steady-state performance of different bypass configurations under load rejection conditions. The computational results indicate that all three bypass control systems exhibit excellent load-following capabilities, effectively responding to a 50 % load step change within 10 s. However, significant differences were observed in their steady-state thermodynamic performance: the upper cycle bypass control achieves the highest steady-state efficiency (η = 29.92 %), followed by the turbine bypass control (η = 27.50 %), with the heat source bypass control showing relatively lower efficiency (η = 26.49 %). Although the upper cycle bypass offers superior efficiency, it leads to a substantial increase in the working fluid flow through the compressor (ΔQ = 15 %) during operation, raising the risk of compressor blockage and necessitating additional flow restriction and anti-blocking interlock control systems. Considering system safety, control complexity, and engineering feasibility, the turbine bypass system, with its relative independence and lower operational risk, is deemed more suitable as the primary control strategy for load rejection conditions.&lt;/div&gt;&lt;div&gt;Building on this, the study proposes a combined control strategy that leverages the strengths of both bypass control and inventory control. During the initial phase of rapid load reduction, the bypass control system quickly adjusts the turbine bypass valve opening to promptly respond to grid load changes, ensuring system frequency stability around 50 Hz, with a maximum deviation of only 0.0193 Hz. Once the load stabilizes, the inventory control system gradually adjusts system pressure and flow rate, reducing bypass flow and ultimately enhancing the system’s thermal efficiency to a new steady-state level. In the four load reduction cases (50 %, 60 %, 70 %, 80 %), the system’s thermal efficiency increases from 27.49 %, 31.25 %, 34.61 %, and 37.63 % to 33.88 %, 36.38 %, 38.43 %, and 40.03 %, respectively. The study also found that even in the 50 % load reduction case, the frequency remains stable around 50 Hz, with a maximum deviation of 0.0193 Hz.&lt;/div&gt;&lt;div&gt;This research provides theoretical and practical guidance for the control strategy of SCO&lt;sub&gt;2&lt;/sub&gt; Brayton cycle systems under rapid load changes, significantly enhancing system safety, efficiency, and reliability, and laying a technical foundation for future wide-load operation in nuclear energy applications.&lt;/div&gt;","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114114"},"PeriodicalIF":1.9,"publicationDate":"2025-05-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143916457","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":"An improved HRA method based on LOOP accident for multi-unit by SPAR-H combined with system dynamics","authors":"Zecong Li,&nbsp;Wanxin Feng,&nbsp;Bingbing Wang,&nbsp;Yu Yu","doi":"10.1016/j.nucengdes.2025.114119","DOIUrl":"10.1016/j.nucengdes.2025.114119","url":null,"abstract":"<div><div>In multi-unit plant sites, existing methods and experiences have identified new issues faced by multi-unit probabilistic safety assessment (PSA), such as the correlation between performance shaping factors (PSFs), the complexity of diagnosis in multi-unit accidents, and the dynamic nature of personnel and environmental states changing over time during human reliability analysis (HRA). Existing methods cannot fully describe the above problems. This paper combines HCR to improve the SPAR-H method, proposes a multi-unit HRA process based on system dynamics (SD), and constructs a dynamic multi-unit HRA model. This model considers the complexity of multi-unit accident diagnosis and establishes the relationship between procedures and time in SPAR-H combined with HCR. SD method is used to represent the relationship between PSFs in SPAR-H, and piecewise functions are used to represent the dynamic changes in personnel and environmental states. In addition, compensation functions for personnel work efficiency using PSFs is added to the model. In this paper, the call of the fifth diesel engine after the emergency diesel engines of both units failed in the loss of off-site power accident is used as a case to demonstrate the feasibility of the proposed model. Based on the calculation results, it can be found that diagnostic errors are the main contributing factor to personnel failure in multi-unit accidents, and the completeness of procedures has a significant impact on personnel errors. The complexity of tasks and the time required for personnel diagnosis can be reduced by supplementing and improving the procedures for multi-unit accidents, thereby reducing the human error probability (HEP) after accidents.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114119"},"PeriodicalIF":1.9,"publicationDate":"2025-05-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143922693","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 start-up performance of a high-temperature sodium heat pipe with a large aspect ratio","authors":"Xu Jinglong ,&nbsp;Men Qiming ,&nbsp;Zhou Quan ,&nbsp;Zhang Zhenhua ,&nbsp;Li Zian ,&nbsp;Qi Baojin ,&nbsp;Wei Jinjia","doi":"10.1016/j.nucengdes.2025.114120","DOIUrl":"10.1016/j.nucengdes.2025.114120","url":null,"abstract":"<div><div>High-temperature sodium heat pipes with a large aspect ratio are widely used for heat dissipation in nuclear reactors. This study investigated high-temperature sodium heat pipes with a large aspect ratio (210) and an ultra-long length (4 m). Hot air was used to preheat the condensation section of the heat pipes to avoid the frozen start-up limit during the starting process of the heat pipes. The experimental results indicated that the start-up time in the horizontal state reached 105 min, and the temperature of the adiabatic section in the preheating stage increased significantly. The condensation section of the heat pipes was then tilted to provide an additional capillary force with the assistance of gravity. Experiments conducted at inclinations of 2° and 5° shortened the start-up time to within 90 min. During the experiments, the sodium heat pipes exhibited the single-point heating lag phenomenon at the front end of the evaporation section. Moreover, the vacuum degree was found to have a significant influence on the starting performance of the heat pipes. Through comparative experiments, this study proved that sufficient wetting time was beneficial to improving the start-up isothermal performance of the heat pipes and prolonging the start-up length. In addition, this study analyzed the reasons for the periodic fluctuations of heat pipes, providing pioneering insights for the stable start-up of high-temperature sodium heat pipes with a large aspect ratio.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114120"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143916423","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":"Subchannel thermal–hydraulic analysis of (UN + U3Si2)-SiC system considering Chalk River Unidentified deposits layer","authors":"Mingdong Kai ,&nbsp;Jiejin Cai","doi":"10.1016/j.nucengdes.2025.114126","DOIUrl":"10.1016/j.nucengdes.2025.114126","url":null,"abstract":"<div><div>The (UN + U₃Si₂)-SiC fuel-cladding combination is one of the most potential accident-tolerant fuel-cladding combinations. In this paper, the influence of surface fouling on core heat transfer is considered, and the thermal–hydraulic characteristics of the accident tolerant fuel (UN + U₃Si₂)-SiC fuel-cladding combination under steady-state conditions and typical LOFA accident conditions are analyzed. Firstly, the model constructed in this paper is systematically validated through PSBT(PWR Subchannel and Bundle Test) benchmark experiments. Then the variation of thermal parameters under steady-state and transient operating conditions when CRUD (Chalk River Unidentified Deposits) is present and at different thicknesses is further explored by analyzing the thermophysical properties of the (UN + U₃Si₂)-SiC fuel-cladding combination. Finally, the safety criterion parameters considering cladding corrosion during LOFA accident are analyzed, and the core safety is judged. The results show that the presence of CRUD will increase the MFCT and MCT of the reactor core and deteriorate the heat transfer. And the thicker the CRUD, the higher the degree of heat transfer deterioration. The existence of CRUD will also reduce the core safe operation parameter MDNBR (Minimal departure from nuclear boiling ratio), but due to the superior performance of the accident tolerant fuel (UN + U₃Si₂)-SiC fuel-cladding combination, the impact of CRUD on the core safe operation is still within the controllable range.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114126"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143916425","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":"Chemical interactions between La(iii) and O2− in molten fluoride salts and La(iii)/U(iv) separation by oxide precipitation method","authors":"Hao Peng ,&nbsp;Nan Ji ,&nbsp;Bo Zhou ,&nbsp;Yunlong Wei ,&nbsp;Wei Huang ,&nbsp;Yu Gong","doi":"10.1016/j.nucengdes.2025.114127","DOIUrl":"10.1016/j.nucengdes.2025.114127","url":null,"abstract":"<div><div>The chemical interactions between La(<span>iii</span>) and O²⁻ in 66.7LiF-33.3BeF₂ (FLiBe) and 46.5LiF-11.5NaF-42KF (FLiNaK) molten salt systems at 873 K were studied by dissolution and oxide titration methods. In the FLiBe, the precipitation-dissolution behavior of La₂O₃ is a simple equilibrium mechanism between La(<span>iii</span>) and O²⁻ ions. The solubility of La₂O₃ in FLiBe melt was 0.078 mol/kg with the dissolution equilibrium time of 5 h, and the corresponding apparent solubility product (<span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span>) of La₂O₃ was (3.43 ± 0.75) × 10⁻⁴ mol⁵/kg⁵. The oxide titration experiment showed that the product of the interaction between La(<span>iii</span>) and O²⁻ in FLiBe is La₂O₃ precipitate, and the <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span> was (3.45 ± 0.37) × 10⁻⁴ mol⁵/kg⁵, which was highly consistent with that obtained by the dissolution method. Based on the <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span> value, the oxide tolerance for La₂O₃ precipitation was then evaluated. However, the chemical reaction between La(<span>iii</span>) and O²⁻ in FLiNaK was more complicated. The dissolution of La₂O₃ would produce oxyfluoride LaOF, and addition of Li₂O into the FLiNaK-La(<span>iii</span>) molten salt could cause precipitation of equimolar solid compounds La₂O₃ and LaOF. The oxyfluoride species LaOF was correlated with a high content of free fluoride ions (F⁻) in FLiNaK. At last, an oxide precipitation method was proposed for La(<span>iii</span>)/U(<span>iv</span>) separation based on the analysis of <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span>(La₂O₃) and <span><math><msubsup><mi>K</mi><mrow><mi>sp</mi></mrow><mo>′</mo></msubsup></math></span>(UO₂), and this method achieved a good La(<span>iii</span>)/U(<span>iv</span>) separation efficiency in the FLiBe-LaF₃-UF₄ melt.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114127"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913236","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":"Material properties and models for the assessment of pressure boundary failure in high-pressure core melt accident scenarios","authors":"Christoph Bläsius ,&nbsp;Jürgen Sievers ,&nbsp;Annett Udoh ,&nbsp;Stefan Weihe","doi":"10.1016/j.nucengdes.2025.114109","DOIUrl":"10.1016/j.nucengdes.2025.114109","url":null,"abstract":"<div><div>The simulation of pressure boundary failure in high-pressure core melt accident scenarios requires specific material data, which are often rare. This work focuses on the behavior of German steel grades and alloys used for KWU-built reactors in this extreme load range. In a first step, relevant materials are described. Previous work is presented in brief. A series of experimental studies addresses open questions regarding the material behavior in specific situations and for specific components, such as short-term creep of Alloy 800 (mod.), tearing behavior of 20 MnMoNi 5 5, relaxation behavior of bolt steels, influence of high-temperature oxidation on fracture, and behavior of contact surfaces in safety and relief valves. Finally, material models for the major steel grades are presented, verified, validated and their accuracy and associated uncertainties are discussed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114109"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913239","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 the spectral element CFD solver on Sunway TaihuLight for nuclear reactor simulation","authors":"Lingyu Dong,&nbsp;Zhifeng Zhou,&nbsp;Genshen Chu,&nbsp;Dandan Chen,&nbsp;Hongzhen Zhang,&nbsp;Yang Li","doi":"10.1016/j.nucengdes.2025.114071","DOIUrl":"10.1016/j.nucengdes.2025.114071","url":null,"abstract":"<div><div>High-fidelity computational fluid dynamics (CFD) plays a crucial role in analyzing thermal–hydraulic phenomena in advanced nuclear reactors. This study presents an optimization of the spectral element method (SEM)-based CFD solver Phiflow-Solver on Sunway TaihuLight supercomputer to accelerate nuclear reactor simulations. The SEM solver relies on small, dense matrix multiplications and the Poisson operator, which are computationally challenging on heterogeneous architectures. To address these challenges, we propose two optimization strategies: (1) Porting matrix operations to the SW26010 processor’s Computing Processing Elements (CPEs) using DMA-enhanced data transfer and SIMD vectorization, achieving a 51.9% performance improvement at 64 CGs for a polynomial order of 24; (2) Enabling collaborative Management Processing Element (MPE)-CPE parallelism to compute multiple spectral elements simultaneously, achieving a 65.5% performance gain under identical conditions. By integrating these strategies, we achieve an overall 70.6% performance enhancement. Validation with a 7-pin wire-wrapped fuel assembly confirms that the heterogeneous optimizations maintain the solver’s accuracy. Furthermore, as the mesh size scales from 42 million to 1.3 billion grid points, the weak scalability remains above 90%, demonstrating the solver’s improved capability for high-resolution nuclear fuel assembly simulations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114071"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143916424","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":"Development of an evaluation method for debris bed formation behavior focusing on the agglomeration mechanism observed in the DEFOR-A test using THERMOS/JBREAK–DPCOOL–MSPREAD","authors":"Wataru Kikuchi ,&nbsp;Akitoshi Hotta ,&nbsp;Koetsu Ito ,&nbsp;Mamoru Shimizu","doi":"10.1016/j.nucengdes.2025.114041","DOIUrl":"10.1016/j.nucengdes.2025.114041","url":null,"abstract":"<div><div>When the lower head of a reactor pressure vessel (RPV) is damaged during a severe accident in light water reactors (LWRs), after the jet breakup occurs in the water, the entrained pieces of molten debris (hereafter called droplets) is likely to form nonuniform, poorly coolable agglomerations on the floor. These debris agglomerates can impact the debris bed coolability. The authors are developing THERMOS, an analysis code composed of modules such as JBREAK, DPCOOL, and MSPREAD, to evaluate these behaviors. In the investigation of the new DEFOR-A test series conducted in collaboration with Kungliga Tekniska Högskolan (KTH), it has been identified that the formation of agglomerated debris is influenced not only by the solidification fraction of the droplets but also by crust cracking and melt spreading. To evaluate the formation of agglomerated debris at a wide range of superheat, the authors have developed the special model in JBREAK, one of the THERMOS modules, based on mechanism estimated from the investigation of DEFOR-A test series (A23-27). Additionally, the agglomeration process is affected by several complex phenomena, such as jet breakup, droplet sedimentation, deposition, and melt spreading behavior, so the authors developed an evaluation method that sequentially evaluates these behaviors using the THERMOS/JBREAK–DPCOOL–MSPREAD coupling. This evaluation method successfully simulated jet breakup, agglomeration, and debris bed formation observed in the DEFOR-A tests. The evaluation method has accurately explained the agglomerated debris mass fraction over a wide range of melt superheat levels by modeling droplet crust cracking, melt spreading, and agglomeration resulting from droplet–debris interactions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114041"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913237","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 progress in high-temperature thermo-mechanical behaviors for modelling Cr-coated cladding under loss-of-coolant accident condition","authors":"Chunyu Yin ,&nbsp;Guanghui Su ,&nbsp;Libo Qian ,&nbsp;Qingwen Xiong ,&nbsp;Yu Liu ,&nbsp;Yingwei Wu ,&nbsp;Sijia Du ,&nbsp;Jing Zhang ,&nbsp;Zhong Xiao","doi":"10.1016/j.nucengdes.2025.114125","DOIUrl":"10.1016/j.nucengdes.2025.114125","url":null,"abstract":"<div><div>Chromium (Cr)-coated zirconium cladding has emerged as a leading candidate for accident tolerant fuel (ATF) cladding in near-term engineering applications. This cladding demonstrates enhanced resistance to high-temperature oxidation, superior mechanical properties at elevated temperatures, and a relatively high level of technological maturity. Its performance under loss-of-coolant accident (LOCA) conditions is critical to reactor safety, making it a key focus of the present study. The present work introduces an overview of research progress on high temperature thermo-mechanical behaviors for Cr-coated cladding and provides a set of fundamental safety analysis models tailored for LOCA scenarios. First, essential models for LOCA safety analysis of Cr-coated cladding are identified, including a high-temperature oxidation model (along with a Cr coating consumption model), a high-temperature creep model, a high-temperature burst model, and an embrittlement criterion. Second, based on the evaluation of experimental data from high-temperature oxidation studies, models for the growth of Cr₂O₃ layer and oxygen absorption are recommended to estimate the oxidation rate of Cr-coated cladding. Additionally, a model for Cr coating consumption is proposed. Subsequently, through a comprehensive review and reevaluation of high-temperature creep and burst data, corresponding models for Cr-coated cladding are developed respectively. Finally, embrittlement data for Cr-coated cladding are analyzed, and embrittlement criteria for both one-sided oxidation and two-sided oxidation conditions are proposed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114125"},"PeriodicalIF":1.9,"publicationDate":"2025-05-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143913238","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}