Nuclear Engineering and Design最新文献

{"title":"Visual experimental research on flow pattern transition mechanisms in vertical helically-coiled tube under liquid-solid-liquid coupled heat transfer conditions","authors":"Hengyuan Wang,&nbsp;Hailin Lei,&nbsp;Qiyu Xuan,&nbsp;Jiacheng Lou,&nbsp;Qisong Yang,&nbsp;Huixiong Li","doi":"10.1016/j.nucengdes.2025.113919","DOIUrl":"10.1016/j.nucengdes.2025.113919","url":null,"abstract":"<div><div>Helically-coiled tube steam generators are important equipment in nuclear power plants. The fluid on the primary side flows over the helically-coiled tube bundles, and exchanges heat with the fluid on the secondary side inside the helically-coiled tube steam generator. Therefore, the secondary side fluid in the helically-coiled tube is under a liquid-solid-liquid coupled heat transfer heating condition. Furthermore, the flow pattern characteristics and transition mechanisms of the secondary side fluid in a helically-coiled tube steam generator play a crucial role in improving the heat transfer efficiency of the generator and the secure and stable operation of nuclear power plants. Therefore, a visual experiment system of a helically-coiled tube steam generator was set up in this paper to observe and study the flow pattern characteristics and transition mechanisms of the secondary side fluid under liquid-solid-liquid coupled heat transfer condition. It was found that the flow patterns of the secondary side fluid include bubble flow, slug flow, stratified-wavy flow and stratified flow in helically-coiled tube. The convergence, contact, collision, coalescence of boiling bubbles and the growth of bubbles along the top of tube were the main flow pattern transition mechanisms under liquid-solid-liquid heat transfer condition. Additionally, the influence of parameters on flow pattern characteristics and transition mechanisms were analyzed, and it was concluded that an increase of heating water temperature in the primary side or an increase of mass flux of secondary side fluid promoted flow pattern transition. However, an increase of mass flux of heating water in the primary side had no obvious effect on the flow pattern transition. Finally, the flow pattern map and the prediction correlations of flow pattern transition boundary curves of the helically-coiled tube were established under liquid-solid-liquid coupled heat transfer conditions based on the experimental results. This flow pattern map was similar to the map obtained by Li (2018), Zhang and Chen (1983), Liu et al. (1976) and Li et al. (2024) based on the water–air experiment in the helically-coiled tube in terms of the shape of some flow pattern transition boundary curve. However, it was significantly different from the flow pattern maps obtained by Mandhane et al., 1974 in a horizontal tube, Mishima and Ishii, 1984 in a vertical tube and Barnea et al. (1980) in a inclined tube.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113919"},"PeriodicalIF":1.9,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143444781","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":"Effective thermal conductivity prediction of dispersion nuclear fuel elements based on deep learning and property-oriented inverse design","authors":"Zekai Huang ,&nbsp;Yingxuan Dong ,&nbsp;Qida Liu ,&nbsp;Xiaoyu Hao ,&nbsp;Hong Zuo ,&nbsp;Qun Li","doi":"10.1016/j.nucengdes.2025.113918","DOIUrl":"10.1016/j.nucengdes.2025.113918","url":null,"abstract":"<div><div>The in-pile thermal performance of dispersion nuclear fuel elements is crucial for reactor safety. The uncertainty of the thermal conduction throughout dispersion fuel is primarily influenced by the nonuniform distribution of fuel particles in the meat, especially the agglomeration behavior of fuel particles. In this paper, a new method has been developed for the rapid and accurate prediction of the effective thermal conductivity (ETC) of dispersion nuclear fuel elements based on the deep learning method. A deep learning model is trained to establish an implicit correlation between the microstructure of dispersion nuclear fuel elements and their ETC, enabling a predictive model to estimate ETC from two-dimensional microstructural diagrams. The dataset is generated using the finite element method, which takes into account the nonuniform distribution characteristic of fuel particles. The microstructures which affect the decision-making of the predictive model are demonstrated by the saliency map. Based on the predictive model, an inverse design method of the distribution of fuel particles was conducted for the specific ETC by metaheuristic algorithms. This study confirms the feasibility of directly evaluating ETC from microstructural images and provides a property-oriented inverse design method for the meat microstructure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113918"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429687","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 control rods position recognition method based capacitance sensing principles within varying liquid level environments","authors":"Yanlin Li,&nbsp;Benke Qin,&nbsp;Hanliang Bo,&nbsp;Wen He","doi":"10.1016/j.nucengdes.2025.113921","DOIUrl":"10.1016/j.nucengdes.2025.113921","url":null,"abstract":"<div><div>The real-time information of the control rods position should be provided by the rod position indicators. Therefore, the rod position indicators (RPIs) are the pivotal safety-related equipment of the nuclear reactors. According to the operating properties of the 200 MW nuclear reactor (NHR200-Ⅱ), the measuring channel of the RPIs fills with two-phase mediums, as the nitrogen and the water. The liquid level of the measuring channel is affected by the operating temperature of the NHR200-Ⅱ. Based on the coaxial type capacitance RPIs, the impact of the rod position and the liquid level on the output of the indicator are studied separately. The position solving model of the control rods is proposed. Based on the model, the dual-channel capacitance RPIs is designed, and the precision of the indicator is evaluated under the cold state and the ordinary pressure. Results indicate that the maximum error of the indicator is within ±4 mm.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113921"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429686","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":"Performance analysis and structural optimization of NaK78-He plate-fin heat exchanger for space nuclear reactor power systems based on a Q3D numerical method","authors":"Liang Yao ,&nbsp;Rui-Tao Liu ,&nbsp;Dan-Dan Su ,&nbsp;Su-Ming Wang ,&nbsp;Lu Wang ,&nbsp;Hong-Na Zhang ,&nbsp;Xiao-Bin Li ,&nbsp;Feng-Chen Li","doi":"10.1016/j.nucengdes.2025.113917","DOIUrl":"10.1016/j.nucengdes.2025.113917","url":null,"abstract":"<div><div>In space nuclear power systems, heat exchangers account for more than 50% of the total system weight. To minimize both weight and volume, this study selected a compact plate-fin heat exchanger (PFHE) as the NaK⁷⁸-He heat exchanger for the space nuclear power system. Computational fluid dynamics (CFD) simulations were employed to analyze the velocity distribution at the NaK⁷⁸ inlet side of a typical PFHE structure, and the channel design was optimized to enhance heat transfer and reduce pressure loss. To investigate the performance variations of the PFHE and further decrease its weight and volume, a simplified and accurate quasi-three-dimensional (Q3D) numerical calculation method was proposed. This method was combined with a genetic algorithm to optimize the PFHE’s size parameters. Results indicated that the uneven flow distribution on the NaK⁷⁸ side could reduce the heat flow by approximately 10% and increase pressure loss by over 500%. The optimized PFHE significantly improved flow uniformity, reducing the standard deviation of velocity distribution from 1.11 to 0.26. Furthermore, after optimization using the genetic algorithm, the PFHE achieved a 23.47% reduction in overall volume and a 22.38% decrease in weight. This study provides valuable insights for optimizing heat exchangers in space nuclear power systems and presents an efficient, rapid, and accurate method for performance evaluation and optimization in engineering applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113917"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429688","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":"Studies on the interaction of sodium with cellulose wool","authors":"M. Venkatesh ,&nbsp;S.Sathish Kumar ,&nbsp;G. Parthiban ,&nbsp;V. Snehalatha ,&nbsp;M. Bootharajan ,&nbsp;M. Muthuganesh ,&nbsp;E. Prabhu ,&nbsp;P.K. Chaurasia ,&nbsp;V.Suresh Kumar ,&nbsp;R. Sudha ,&nbsp;E.Hemanth Rao ,&nbsp;K. Sundararajan ,&nbsp;Sanjay Kumar Das ,&nbsp;Rajesh Ganesan ,&nbsp;V. Jayaraman ,&nbsp;D. Ponraju ,&nbsp;N. Sivaraman","doi":"10.1016/j.nucengdes.2025.113910","DOIUrl":"10.1016/j.nucengdes.2025.113910","url":null,"abstract":"<div><div>This study investigates the interaction between liquid sodium and cellulose wool at various temperatures, commonly used in fast breeder reactors (FBR). Cellulose wool, used for cleaning reactor components, was subjected to sodium interaction experiments within the welded stainless steel capsules in argon heated at temperatures from 150 °C to 550 °C. Analysis was carried out for the gaseous reaction products and solid residue using various characterization techniques. The products showed the presence of H₂, CO, CO₂, CH₄, C₂H₂, C₂H₄, C₂H₆ and residual carbon. Gaseous products from the reactions were analyzed using Gas Chromatography (GC) with flame ionisation detector (GC-FID), Discharge ionization detector (GC-DID) and metal oxide semiconducting sensors. The gaseous products at 150 °C indicated the presence of CO and CO₂ as the sample temperature was increased these gases were absent. The concentration of H₂, C₂H₄ and C₂H₆ in the gaseous product decreased while the CH₄ concentration increased with temperature. Solid residues were analysed by (i) Scanning electron microscope/Energy dispersive (SEM/EDS) for morphology and chemical content, (ii) carbon analysis and (iii) infrared spectroscopy (IR). The systematic study using these techniques revealed a significant decomposition of cellulose wool with liquid sodium at high temperatures.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113910"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429684","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 and application of safety analysis code for He-Xe cooled space reactor","authors":"Ming Liu ,&nbsp;Pan Wu ,&nbsp;Ke Huang ,&nbsp;Jianqiang Shan","doi":"10.1016/j.nucengdes.2025.113914","DOIUrl":"10.1016/j.nucengdes.2025.113914","url":null,"abstract":"<div><div>Space nuclear reactor power system can offer high energy density and a wide power coverage, providing electric power for space exploration. The safety characteristics of space nuclear reactor during in-orbit operation accidents is key to evaluate its feasibility. In this paper, a safety analysis code is developed for He-Xe mixture cooled space reactor based on SCTRAN, which used to be applied for Supercritical water or CO₂ cooled reactor. A semi empirical formula is applied to develop a physical property calculation model of He–Xe. Flow resistance and heat transfer correlations as well as turbomachinery models for He-Xe are developed. A numerical model for S⁴-CBC space reactor power system is established by using the newly developed code. The calculation results of reactor and system model have the maximum relative errors of 3.43 % and 0.47 % compared to the reference design values, which proves the code is reliable to simulate the thermal–hydraulic behavior of the S⁴-CBC space reactor power system. The transient responses of the S⁴-CBC space reactor power system under the reactivity insertion accident and the mechanical failure of one and two Brayton loops are simulated and analyzed. The results of the accident analysis provide guidance for the core design, operation strategy, and safety system design of future He-Xe cooled space reactor system.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113914"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429685","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":"Behavior mechanism and prediction models of stress corrosion cracking of stainless steels in pressurized water reactors","authors":"Pengfei Gao ,&nbsp;Yanhui Li ,&nbsp;Zhouyang Bai ,&nbsp;Shaoming Ding ,&nbsp;Yinan Zhang ,&nbsp;Limei Xing ,&nbsp;Zhihong Yu","doi":"10.1016/j.nucengdes.2025.113897","DOIUrl":"10.1016/j.nucengdes.2025.113897","url":null,"abstract":"<div><div>Stress corrosion cracking (SCC) which has long been an essential topic for alloys exposed to subcritical water systems,is an urgent issue for pressurized water reactors (PWRs). Due to the complexity of the interaction of materials, environment, and stress, the SCC mechanism of consensus and the prediction model with mechanic extensibility are unavailable. Based on reviewing the typical characteristics and critical influence factors of SCC of stainless steels, this paper systematically summarizes, compares the more mainstream cracking mechanisms, and from the perspective of empirical and deterministic methods, discusses the applicable conditions of a series of SCC prediction models with high recognition via practical data, which are of great significance for strengthening readers’ understanding of SCC critical parameters and the matching service environment of each SCC mechanism, promoting the accurate prediction of SCC behavior and cumulative damage in the pressurized water reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113897"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429689","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":"Interaction characteristics of subcooled water jet with high-pressure upward into high-temperature oil in a confined space","authors":"Shuhua Zhou,&nbsp;Ruizhi Hao,&nbsp;Wenshu Jiang,&nbsp;Shuo Wang,&nbsp;Hanwu Gao,&nbsp;Xinkui Fang,&nbsp;Kuo Bian,&nbsp;Xue Chen,&nbsp;Tao Lu","doi":"10.1016/j.nucengdes.2025.113913","DOIUrl":"10.1016/j.nucengdes.2025.113913","url":null,"abstract":"<div><div>During the Steam Generator Tube Rupture (SGTR) accident of Lead-cooled Fast Reactors (LFR), liquid metals come into direct contact with water, resulting in a multiphase flow phenomenon involving liquid metal, water and vapor. Studying the corresponding mechanisms of the interaction process between subcooled water jet and liquid metal is crucial with the pressure impact and temperature drastic changes during the phenomenon may damage the structural components of the reactors. The simulation experiments were conducted to investigate the interaction characteristics of subcooled water jet with high-pressure upward into high-temperature thermal oil in a confined space, studying the effect of experimental parameters (such as injection time and pressure, oil and water temperature, nozzle shape and diameter) on the oil temperature drop and pressure variation. Three pressurization modes were distinguished, including two overall vapor productions and one local vapor production. The evolution path of water/vapor in oil was inferred based on transient temperature changes. Finally, the oil temperature drop model was established by four dimensionless thermophysical parameters, and the model was validated with experimental values at an average error of 12.05%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113913"},"PeriodicalIF":1.9,"publicationDate":"2025-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421987","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 failure criterion for nuclear fuel cladding due to internal gas","authors":"Mingda Han ,&nbsp;Hetong Liu ,&nbsp;Weixu Zhang ,&nbsp;Yibo Zhang ,&nbsp;Songhui Luo","doi":"10.1016/j.nucengdes.2025.113909","DOIUrl":"10.1016/j.nucengdes.2025.113909","url":null,"abstract":"<div><div>The primary method for assessing cladding failure due to internal gas pressure involves comparing peak and ultimate pressures. When a small amount of gas is present, the pressure decreases as the volume expands at high temperatures, allowing the cladding to operate safely. This paper proposes a gas equilibrium-based failure criterion for cladding, determined by comparing the sectional force required by the gas with what the structure can provide. Finite element analysis compared cladding failure under three different boundary conditions (most used), using bidirectional fluid–structure interaction to investigate failure from internal gas. The critical load-bearing gas moles and the cross-sectional force changes with radial displacement for both hollow cladding and cladding with a fuel pellet were investigated. It was found that the critical gas quantity ratio is significantly smaller than the volume ratio, with gas pressure decreasing more rapidly in the pellet case.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113909"},"PeriodicalIF":1.9,"publicationDate":"2025-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421988","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 studies on nitrogen’s effect on reactor core cooling during a hot leg SBLOCA in a scaled EPR model","authors":"Vesa Riikonen,&nbsp;Virpi Kouhia,&nbsp;Markku Puustinen,&nbsp;Giteshkumar Patel,&nbsp;Antti Räsänen,&nbsp;Eetu Kotro,&nbsp;Juhani Hyvärinen","doi":"10.1016/j.nucengdes.2025.113916","DOIUrl":"10.1016/j.nucengdes.2025.113916","url":null,"abstract":"<div><div>The presence of non-condensable gases in the reactor cooling system can significantly influence the operation of several safety systems within a nuclear power plant. The pressurized nitrogen volume at the top of the accumulator tank is the driving force for injecting accumulator water. In certain nuclear power plants, the release of gaseous nitrogen to the primary side is inhibited by an automatic closure of the accumulator injection line at the end of the discharge. However, if this automatic closure system fails, nitrogen will inadvertently flow into the reactor cooling system once the accumulators have been depleted. Furthermore, it is important also to note that the water within the accumulator is saturated with dissolved nitrogen, resulting in the injection of some nitrogen into the primary system alongside every accumulator discharge.</div><div>The impact of nitrogen on core cooling during loss of coolant accidents (LOCAs) has been investigated experimentally using the PWR PACTEL facility. The main observations were that when a break occurs in the hot leg, the injection of nitrogen from an accumulator can effectively prevent depressurization of the primary side. Consequently, the core can experience a heat-up at primary pressure somewhat above the typical low-pressure safety injection (LPSI) shut-off head. The decoupling of primary and secondary side pressures depends on the amount of nitrogen released from the accumulator, how much of it accumulates into the U-shaped steam generator heat exchange tubes, decreasing the condensation heat transfer, and the number of steam generators participating in the secondary side depressurization. Furthermore, the size of the break significantly affects the volume of nitrogen escaping through the break, which in turn influences the nitrogen levels within the system. Specifically, larger breaks permit a greater flow of nitrogen, thereby reducing the likelihood of disrupting heat transfer between the primary and secondary sides while also mitigating the depressurization of the primary side.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113916"},"PeriodicalIF":1.9,"publicationDate":"2025-02-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421985","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}