International Journal of Advanced Nuclear Reactor Design and Technology最新文献_第4页

Research on modeling and fuel performance assessing of silicon-modified ferritic-martensitic steel cladding in flowing lead–bismuth eutectic corrosion environment 流动铅铋共晶腐蚀环境下硅改性铁素体-马氏体钢包层建模及燃料性能评估研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-12-01 Epub Date: 2025-10-13 DOI: 10.1016/j.jandt.2025.10.002

Runjie Fang, Guoliang Zhang, Yuyu Lin, Bo Zhang, Shaofang Lin, Qisen Ren

{"title":"Research on modeling and fuel performance assessing of silicon-modified ferritic-martensitic steel cladding in flowing lead–bismuth eutectic corrosion environment","authors":"Runjie Fang, Guoliang Zhang, Yuyu Lin, Bo Zhang, Shaofang Lin, Qisen Ren","doi":"10.1016/j.jandt.2025.10.002","DOIUrl":"10.1016/j.jandt.2025.10.002","url":null,"abstract":"<div><div>Lead-bismuth coolant corrosion causes thinning of stainless steel fuel rod cladding through oxidation, dissolution, and erosion effects, thereby weakens fuel rod mechanical performance and degrades heat transfer due to its low thermal conductivity. The application of ferritic-martensitic stainless steel (F/M steel) modified with silicon can further enhance its resistance to lead-bismuth corrosion, making it an ideal cladding material choice for advanced lead-bismuth reactors. To analyze the behavior of F/M steel fuel rods under lead-bismuth corrosion conditions with higher accuracy, a silicon-modified F/M steel cladding corrosion model applicable to flowing oxygen-controlled lead-bismuth coolant environments based on previous public data is developed and subjected to preliminary validation and uncertainty analysis. Subsequently, fuel rod performance analyses are conducted under typical operating parameters of lead-bismuth reactors. The results demonstrate that silicon modified F/M steel cladding fuel rods can withstand long-term high-temperature neutron irradiation and lead-bismuth corrosion. Both steady and transient states conditions satisfy all design criteria with significant margins. Even with the upper envelope corrosion model and comparing with normal F/M steel, the calculations confirm that silicon modified F/M steel materials maintain significant corrosion resistance and high-temperature mechanical stability, capable of adapting to the operating requirements of lead-bismuth reactors under extreme conditions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 468-475"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145467088","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Thermo-mechanical modeling of residual stress and filler metal optimization for SiC cladding joints SiC熔覆接头残余应力热力学建模及填充金属优化

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-12-01 Epub Date: 2025-11-26 DOI: 10.1016/j.jandt.2025.11.004

Yongheng Lu , Yu Zhou , Shuang Liang , Zaihong Wang , Yue Jia , Chong Wei

{"title":"Thermo-mechanical modeling of residual stress and filler metal optimization for SiC cladding joints","authors":"Yongheng Lu , Yu Zhou , Shuang Liang , Zaihong Wang , Yue Jia , Chong Wei","doi":"10.1016/j.jandt.2025.11.004","DOIUrl":"10.1016/j.jandt.2025.11.004","url":null,"abstract":"<div><div>This study employed the finite element modeling method to investigate the thermodynamic behavior of brazed joints between silicon carbide (SiC) clad tubes and five metal-based filled alloys (Ti, Ni, Zr, Ta, Nb). A thermally mechanical coupling model was established to simulate the thermal stress and strain distribution within the temperature range of 20–1200 °C. The research results provide clear criteria for the selection of filled alloys under different temperature conditions. Below 800 °C, due to the low coefficient of thermal expansion (CTE) and elastic modulus of ZR-based fillers, they have good compatibility with SiC and the lowest residual stress (63.38 MPa at 360 °C and 109.34 MPa at 800 °C). At 1200 °C, niobium-based fillers exhibit the best performance, with a residual stress of 1189.55 mpa, which is approximately 50 % lower than that of Ta-based fillers. Moreover, due to its high melting point, it has stable plastic deformation resistance. However, Ni-based solder generates extremely high stress (1848.68 MPa) at 800 °C, far exceeding its yield strength, and has a high risk of interface failure. Ti-based fillers exhibit moderate stress relaxation through plastic deformation near 600 °C, but they soften significantly at higher temperatures. Finite element analysis clarified the potential mechanism of residual stress generation caused by CTE mismatch, expanding the discussion space for the hierarchical design of high-integrity silicon carbide clad joints.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 494-504"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145617977","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Development and application of an engineering simulator for an integrated heating reactor 一体化加热堆工程模拟器的研制与应用

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-12-01 Epub Date: 2025-11-24 DOI: 10.1016/j.jandt.2025.11.002

Yuanlei He , Xujia Wang , Jun Sun , Shuhan Yang , Wei Zhou

{"title":"Development and application of an engineering simulator for an integrated heating reactor","authors":"Yuanlei He , Xujia Wang , Jun Sun , Shuhan Yang , Wei Zhou","doi":"10.1016/j.jandt.2025.11.002","DOIUrl":"10.1016/j.jandt.2025.11.002","url":null,"abstract":"<div><div>The integrated heating reactor is a novel design intended to meet diverse energy demands, including industrial steam supply, district heating, and seawater desalination. To address key challenges in its development—such as verifying control strategies, conducting closed-loop validation of processes, instrumentation and control (I&C), and human–machine interface systems, and ensuring the correctness and practicality of operating procedures—a dedicated engineering simulator is essential. Building on the reactor's specific design features, this study analyzes key enabling technologies, develops an engineering simulator, and uses it to verify and optimize operational control schemes. The investigation reveals an inherent inertia-induced delay in power ramping unique to fully natural-circulation reactors. By applying optimized control system, oscillations in reactor power, primary loop average temperature, and secondary loop steam pressure during ramp power change transients are effectively suppressed. The developed engineering simulator therefore serves as a powerful platform for evaluating control strategies, providing crucial support for closed-loop verification of processes and instrumentation, and refining operational procedures.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 486-493"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145617978","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Effects of turbulent heat diffusion and viscosity on the simulation of turbulent mixing in autonomous PWR subchannel analysis software 自主压水堆子通道分析软件中湍流热扩散和粘度对湍流混合模拟的影响

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-12-01 Epub Date: 2025-11-19 DOI: 10.1016/j.jandt.2025.11.001

Bin Jia, Jie-Ru An, Guang-Yu Zhu, Xin-Li Gao, Xue-Dong Qiao, Xiao-Qiang He

{"title":"Effects of turbulent heat diffusion and viscosity on the simulation of turbulent mixing in autonomous PWR subchannel analysis software","authors":"Bin Jia, Jie-Ru An, Guang-Yu Zhu, Xin-Li Gao, Xue-Dong Qiao, Xiao-Qiang He","doi":"10.1016/j.jandt.2025.11.001","DOIUrl":"10.1016/j.jandt.2025.11.001","url":null,"abstract":"<div><div>In this paper the turbulent mixing models of several domestic autonomous PWR subchannel software under review by the national nuclear safety administration were evaluated and studied. Firstly, it is found that most of these subchannel software adopt the equal-mass turbulent mixing model. Then, an experimental validation study was performed on the equal-mass turbulent mixing model. Significant variations in the model's computational coefficients were observed under different operating conditions. Finally, parameter sensitivity analysis and engineering validation were conducted. It is obtained: In single-phase conditions, energy transfer between hot and cold channels is insignificant, while momentum transfer dominates. Effective simulation of turbulent mixing is achieved by adjusting the turbulent dynamic viscosity coefficient M<sub>t</sub>. In two-phase conditions, energy transfer between hot and cold channels dominates, while the influence of momentum transfer is negligible. Effective simulation of turbulent mixing is achieved by adjusting the turbulent thermal diffusion coefficient K<sub>t</sub>. In actual PWR nuclear power plant operations, the core condition is primarily a two-phase state with subcooled boiling. Therefore, for the current subchannel software, turbulent mixing is effectively simulated mainly by adjusting the turbulent thermal diffusion coefficient K<sub>t</sub>, while the influence of turbulent dynamic viscosity is generally neglected.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 511-518"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145736103","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Chemical speciation analysis of fission products in the PWR fuel pellet during accidents 事故中压水堆燃料球团裂变产物的化学形态分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-12-01 Epub Date: 2025-11-24 DOI: 10.1016/j.jandt.2025.11.003

Wang Zhenghui, Lu Changdong

{"title":"Chemical speciation analysis of fission products in the PWR fuel pellet during accidents","authors":"Wang Zhenghui, Lu Changdong","doi":"10.1016/j.jandt.2025.11.003","DOIUrl":"10.1016/j.jandt.2025.11.003","url":null,"abstract":"<div><div>During the operation of a pressurized water reactor, large amounts of fission products (FP) are generated in the grains of fuel pellets. The transport behavior of different chemical species of fission products varies under nuclear power plant accidents. In this paper, the diffusion and release characteristics of fission products in UO<sub>2</sub> grains are obtained through a comprehensive chemical thermodynamic analysis of all phase states of the fission products within the fuel pellets. The diffusion of fission gases follows Fick's law and is calculated using the Booth model, with the release rate related to the diffusion coefficient. The release mechanism of volatile fission products is similar to that of fission gases, sharing the same diffusion coefficient. The release of semi-volatile fission products requires consideration of the proportion of gaseous products, with the gaseous forms of these fission products within the pellets determined by their chemical thermodynamic properties. Therefore, the diffusion coefficients of fission products (excluding krypton and xenon) can be obtained by multiplying the diffusion coefficient of fission gases by an appropriate scaling factor <em>Sk</em> for each radionuclide. This paper determines the <em>Sk</em> values for various elements through chemical thermodynamic analysis, which can be used for the analysis of fission product release under accident conditions.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 505-510"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145684303","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Proposal of an online univariate monitoring algorithm for sensors in primary circuit of nuclear reactor based on sliding-window kernel density estimation and quantitative metrics 提出了一种基于滑动窗口核密度估计和定量度量的核反应堆一次电路传感器在线单变量监测算法

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-12-01 Epub Date: 2025-10-10 DOI: 10.1016/j.jandt.2025.10.001

Fenqin Xu , Xiaoyu Yan , Yangbin Deng , Jiantao Zheng , Bo Pang , Dou Zhao , Yan Tu , Zhan Xing

{"title":"Proposal of an online univariate monitoring algorithm for sensors in primary circuit of nuclear reactor based on sliding-window kernel density estimation and quantitative metrics","authors":"Fenqin Xu , Xiaoyu Yan , Yangbin Deng , Jiantao Zheng , Bo Pang , Dou Zhao , Yan Tu , Zhan Xing","doi":"10.1016/j.jandt.2025.10.001","DOIUrl":"10.1016/j.jandt.2025.10.001","url":null,"abstract":"<div><div>The reliability and accuracy of safety-related sensors in the primary circuit of a nuclear reactor are crucial to ensure the safety and economic viability of power plants. In recent years, substantial research has focused on the development of online monitoring (OLM) and condition-based maintenance (CBM) methods to enhance conventional maintenance strategies that rely on periodic manual calibration (PMC). However, certain measurements from the power plant instrumentation system exhibit weak correlations with other variables. These uncorrelated signals are categorized as either independent or isolated signals. Although these signals are unsuitable for inclusion in multivariate-based OLM systems, they still require monitoring. This study introduces a novel data-driven algorithm based on sliding-window kernel density estimation (SWKDE) combined with quantitative metrics for online statistical univariate monitoring. The effectiveness of the proposed method was demonstrated by detecting sensor faults using data retrieved from the CPR1000 full-scope pressurized water reactor (PWR) simulator. Results of the case studies indicated that typical types of sensor soft faults, including bias, drift, and noise faults, can be successfully detected by identifying data anomaly patterns using qualitative metrics.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 4","pages":"Pages 405-414"},"PeriodicalIF":0.0,"publicationDate":"2025-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145321784","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Study on the method of measuring the interface of radioactive liquid by air blowing 吹气法测量放射性液体界面的方法研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-09-01 Epub Date: 2025-06-08 DOI: 10.1016/j.jandt.2025.06.003

Fengna Tang , Jianwei Liu , Daoliang Chen

引用次数: 0

Impact of revised RG 1.183 on source terms and radiological consequences of rod ejection accident for the HRP1000 reactor 修订后的RG 1.183对HRP1000反应堆燃料棒抛射事故源项和辐射后果的影响

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-09-01 Epub Date: 2025-06-19 DOI: 10.1016/j.jandt.2025.06.007

Yichen Chen, Changdong Lu, Zewu Zhao, Pingting Jiang, Chao Yu

{"title":"Impact of revised RG 1.183 on source terms and radiological consequences of rod ejection accident for the HRP1000 reactor","authors":"Yichen Chen, Changdong Lu, Zewu Zhao, Pingting Jiang, Chao Yu","doi":"10.1016/j.jandt.2025.06.007","DOIUrl":"10.1016/j.jandt.2025.06.007","url":null,"abstract":"<div><div>This paper compares and analyzes the impact of the revised RG 1.183 on the source term and radioactive consequences of the rod ejection accident. The revision of RG 1.183 has resulted in changes to key assumptions in the source term analysis of the rod ejection accident, including the release fractions of fission products from the fuel rod plenum and gap, the release fractions from the molten fuel rod, as well as the primary-to-secondary leakage rates of the SG. The analysis shows that, based on the typical radial average fuel enthalpy increase after a rod ejection accident in the HRP1000, the revised RG 1.183 leads to a slight increase in the dose consequences for the public at the site boundary via the containment release pathway, while the dose consequences via the secondary system release pathway are slightly reduced. Additionally, the dose consequences of the rod ejection accident are positively correlated with the radial average fuel enthalpy increase. However, even if the radial average fuel enthalpy increase rises to 70 cal/g, the radioactive consequences still meet the acceptance criteria for design basis accidents as specified in GB 6249–2025.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 3","pages":"Pages 250-256"},"PeriodicalIF":0.0,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144523112","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Simulation study of steam flow behavior and piping mechanical response analysis in safety valve discharge pipes with fluid-structure interaction 流固耦合安全阀排放管道蒸汽流动特性仿真研究及管道力学响应分析

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-09-01 Epub Date: 2025-07-29 DOI: 10.1016/j.jandt.2025.07.006

Shasha Yin , Xietian Ji , Xiaodong Lu , Qi Zhang

{"title":"Simulation study of steam flow behavior and piping mechanical response analysis in safety valve discharge pipes with fluid-structure interaction","authors":"Shasha Yin , Xietian Ji , Xiaodong Lu , Qi Zhang","doi":"10.1016/j.jandt.2025.07.006","DOIUrl":"10.1016/j.jandt.2025.07.006","url":null,"abstract":"<div><div>This study numerically simulates the steam flow behavior and piping mechanical response within the safety valve discharge piping of a nuclear power system. A one-way Fluid-Structure Interaction (FSI) approach was employed to simulate high-temperature saturated steam flow through an adiabatic pipe featuring three 90° right-angle bends. Model reliability was ensured via mesh independence verification and comparison with theoretical solutions. Analysis of steam flow parameter distribution and pipe stress-strain characteristics under mass flow inlet and pressure inlet boundary conditions reveals that centrifugal forces and Dean vortices induce significant pressure/velocity stratification within the elbow region. A low-pressure stagnation zone along the inner wall coexists with a high-velocity zone on the outer wall, amplifying localized energy dissipation. Under low-pressure conditions (0.8 MPa), steam compressibility triggers flow transition, generating supersonic flow (<em>Ma</em> = 1.77) at the outer elbow. Boundary conditions critically govern mechanical responses, with pressure inlet conditions causing abrupt fluid kinetic energy surges that elevate elbow impact stress to 430.39 MPa-surpassing the material yield limit. The second elbow emerges as a structural vulnerability due to cumulative upstream vortices, exhibiting maximum strain of 12.09 mm.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 3","pages":"Pages 308-318"},"PeriodicalIF":0.0,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738825","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Research on temperature composite control system for radioactive waste incineration equipment 放射性废物焚烧设备温度复合控制系统的研究

International Journal of Advanced Nuclear Reactor Design and Technology Pub Date : 2025-09-01 Epub Date: 2025-06-06 DOI: 10.1016/j.jandt.2025.06.004

Zhiyong Wang , Yao Cui , Jing Ma , Bo Zhang , Liudong Hou , Haili Li

{"title":"Research on temperature composite control system for radioactive waste incineration equipment","authors":"Zhiyong Wang , Yao Cui , Jing Ma , Bo Zhang , Liudong Hou , Haili Li","doi":"10.1016/j.jandt.2025.06.004","DOIUrl":"10.1016/j.jandt.2025.06.004","url":null,"abstract":"<div><div>This paper presents a multi-mode control scheme for the temperature regulation of radioactive waste incineration equipment, integrating modern and fuzzy control theories. Separate controllers were designed for the preheating and operating stages. In preheating phase, a mathematical model was developed using mechanism modeling, enabling automatic control through sequential and proportional-integral-derivative control. For operation phase, a composite control approach combining fuzzy control and special variable ratio control was proposed. The fuzzy controller used temperature difference, changing rate and acceleration as inputs, adjusting oxygen concentration to regulate combustion temperature. Experimental results demonstrated 27 h of stable automatic control with a maximum temperature deviation of 9.1 °C and an error rate of 1.65 %. The proposed scheme provides specific control strategies with quantifiable performance indicators, proving its effectiveness in maintaining temperature stability in radioactive waste incineration.</div></div>","PeriodicalId":100689,"journal":{"name":"International Journal of Advanced Nuclear Reactor Design and Technology","volume":"7 3","pages":"Pages 209-216"},"PeriodicalIF":0.0,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144480525","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0