Nuclear Engineering and Design最新文献

{"title":"Axial-flow-induced structural vibration in a 5×5 cylinder cluster","authors":"Zongyan Lu ,&nbsp;Peng Wang ,&nbsp;Yu Zhou","doi":"10.1016/j.nucengdes.2024.113739","DOIUrl":"10.1016/j.nucengdes.2024.113739","url":null,"abstract":"<div><div>This work aims to experimentally study the incident turbulence intensity <em>T_u</em> effect on the flow-induced vibration of an elastic cylinder positioned at the center of a 9- or 25-cylinder cluster subjected to an axial flow. <em>T_u</em> is examined at 0.71% – 0.80% and 2.30% – 2.91%. The pitch-to-diameter ratio <em>P*</em> is 1.36 ∼ 1.64. Lateral vibrations along two orthogonal directions are simultaneously measured with the interstitial flow of the cylinder bundle. Two mechanisms are identified behind the elastic-cylinder vibration at low and high <em>T_u</em>. One is the presence of a varying velocity gradient within the cylinder bundle, and the other is incident flow fluctuations. At low <em>T_u</em> (0.71% – 0.80%), the root-mean-square vibration amplitude <em>A_rms*</em> of the elastic cylinder exhibits strong dependence on the <em>P*</em> and cylinder number <em>N</em>. Increasing velocity gradient with decreasing <em>P*</em> or increasing <em>N</em> plays a key role in destabilizing the shear layers surrounding the elastic cylinder, inducing eddies to separate from the cylinder-wall and actively interact with those near the neighboring cylinder. Therefore, the near-wall velocity fluctuation <em>u_rms*</em> and <em>A_rms*</em> are increased. At high <em>T_u</em> (2.3% – 2.91%), <em>A_rms*</em> is weakly dependent on <em>P*</em> compared with that at low <em>T_u</em>. It is found that the shear-layer instability surrounding the elastic cylinder is mainly intensified by the incident flow fluctuations with a higher <em>T_u</em>, accounting for the enhanced eddy activities, while the velocity-gradient effect associated with a change in <em>P*</em> is of less importance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113739"},"PeriodicalIF":1.9,"publicationDate":"2024-12-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142757644","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 improvement of the generator end fixed structure and optimization of pretightening process","authors":"Zhifu Tan,&nbsp;Lidong He,&nbsp;Chunyan Deng,&nbsp;Xingyun Jia,&nbsp;Gang Zhu","doi":"10.1016/j.nucengdes.2024.113747","DOIUrl":"10.1016/j.nucengdes.2024.113747","url":null,"abstract":"<div><div>To address recurring issues identified during comprehensive maintenance of a specific generator model—particularly the loosening of ring lead fixing structures at the generator ends, along with the detachment of felt and insulation wear—a detailed investigation was conducted. This study focused on analyzing the root causes of these problems and developing targeted improvement strategies. Taking the ring lead fixing structure as the research object, the investigation explored enhancements to the fixing structure and optimization of the bolt pretightening process. By comparing the fixing structure of faulty units with that of fault-free units and considering practical field conditions, a solution involving an increased cleat wrapping angle was proposed. The effectiveness of this improved design was verified through testing. Furthermore, the interaction of adjacent bolts during the stretching of single-head bolt stretchers was investigated for the bolt pre-tightening process. In view of the bolt pretightening process, the study explored the interaction between adjacent bolts when utilizing single-head bolt tensioners was studied, and experiments using multi-head synchronous hydraulic tensioners were conducted to evaluate their pretightening performance. The results revealed the pretightening behavior of multi-head systems, demonstrating their superior efficiency compared to single-head methods. These findings provide a robust foundation for applying such systems in industrial settings. This study aims to resolve technical challenges related to the ring lead fixing structure and the bolt pretightening process, ultimately ensuring the safe and stable operation of the generator.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113747"},"PeriodicalIF":1.9,"publicationDate":"2024-11-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142744066","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":"Mechanical performance and self-sensing capability of an amorphous fiber-reinforced concrete for radioactive waste storage facilities","authors":"Théophile Bouillard ,&nbsp;Anaclet Turatsinze ,&nbsp;Olivier Helson ,&nbsp;Jean-Paul Balayssac ,&nbsp;Ahmed Toumi ,&nbsp;Carole Soula","doi":"10.1016/j.nucengdes.2024.113727","DOIUrl":"10.1016/j.nucengdes.2024.113727","url":null,"abstract":"<div><div>This study results from a research project initiated by the French National Radioactive Waste Management Agency (Andra) as a part of the Cigéo (Industrial Center for Geological Disposal) project. This project addresses many challenges related to sustainability, and health monitoring of confinement structures planned by Andra. In order to reduce issues related to corrosion in the tunnels, fiber reinforcement can be used to reduce the amount of steel reinforcement while maintaining high mechanical performance and durability. The fibers used in this study are named Fibraflex, metallic fibers provided by Saint Gobain SEVA. Various aspect ratios (82 and 123) and volume ratios (0.27% and 0.41%) of fibers were tested. These fibers feature high specific surface area (around 70 m²/m³), corrosion resistance, and high electrical conductivity. To characterize the contribution of fibers to concrete mechanical performances, European standard EN 14651 flexural tensile tests were conducted. Results show that fibers bring better performances, especially concerning the post-peak behavior. In fact, fibers effectively transfer stresses across cracks, resulting in better residual tensile strengths. Since these fibers are electrically conductive, tests were also conducted to design self-sensing concrete based on electrical measurements. Firstly, resistivity was monitored on saturated specimens with different mix design throughout the curing time. Results show that when fibers were added, concrete resistivity was reduced due to the ability of the fibers to transmit electrons. A relationship between reinforcement index and resistivity was thus revealed. The self-sensing capacity was investigated on specimens subjected to cyclic bending loading, with electrical properties measured by using a Wheatstone bridge. Results show a good agreement between the level of cracking in concrete and its electrical response.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113727"},"PeriodicalIF":1.9,"publicationDate":"2024-11-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142744067","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":"Small break LOCA studies for different layouts of passive safety systems in the IRIS reactor","authors":"Siniša Šadek,&nbsp;Davor Grgić,&nbsp;Paulina Družijanić","doi":"10.1016/j.nucengdes.2024.113745","DOIUrl":"10.1016/j.nucengdes.2024.113745","url":null,"abstract":"<div><div>IRIS (International Reactor Innovative and Secure) is an integral, medium power, light water reactor with advanced safety features. In the first decade of the 21^st century, 22 institutions under the leadership of Westinghouse Electric Corporation were involved in its development. The University of Zagreb, along with the Polytechnic of Milan, was in charge of performing safety analyses. A detailed plant model is developed using the RELAP5 code for the analyses of thermal–hydraulic processes in the reactor vessel, the GOTHIC code for the analysis of the processes in the containment and, in addition, the ASYST code for the calculation of a severe accident. Some of the previous small break loss-of-coolant accident analyzes at the existing pipelines are repeated to test the improved plant model. However, the focus of the paper is on the new set of analyzes of hypothetical breaks along the reactor vessel with the aim of determining whether the passive safety systems can ensure successful core cooling. For this purpose, two models are developed with different configurations of the emergency heat removal system and the safety systems inside the containment that inject water into the reactor vessel. The results show the complex and rather ambiguous dependence of the reactor coolant system thermal–hydraulic behaviour on the selected boundary conditions. The scenarios analyzed vary from design basis events to severe accidents. The capabilities of specific safety systems in mitigating the consequences of an accident are determined, depending on the position and size of the break on the reactor vessel wall.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113745"},"PeriodicalIF":1.9,"publicationDate":"2024-11-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142757643","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":"Detection of flow instability in natural circulation boiling water reactors using instantaneous spectral entropy measure","authors":"Ch Santosh Subudhi,&nbsp;Sreyas Rajagopal Shimjith","doi":"10.1016/j.nucengdes.2024.113675","DOIUrl":"10.1016/j.nucengdes.2024.113675","url":null,"abstract":"<div><div>Natural Circulation Boiling Water Reactors (NCBWR) are vulnerable to thermal-hydraulic instabilities in coolant flow under certain operating conditions. Considering the strong coupling between neutronics and thermal hydraulics, monitoring of such instabilities is crucial for safe and reliable operation of NCBWRs. However, wider frequency range, significant non-stationarity and time-varying nature of flow oscillations makes this a challenging task. In this paper, we introduce a novel method using Instantaneous Spectral Entropy (ISE) to detect and evaluate flow instabilities. Key advantage of the method is its ability to clearly differentiate between random noise and oscillations. Comparison between Short-Time Fourier Transform (STFT) and Continuous Wavelet Transform (CWT) for calculating entropy revealed that STFT provides superior results. Validation with operational data from a thermal hydraulic facility confirmed the method’s effectiveness in identifying and assessing time-varying flow instabilities.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113675"},"PeriodicalIF":1.9,"publicationDate":"2024-11-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142757732","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 potential burnable poison materials for a small modular block-type HTGR design using MgO-BeO as a composite-based moderators","authors":"Irwan L. Simanullang,&nbsp;Nozomu Fujimoto","doi":"10.1016/j.nucengdes.2024.113742","DOIUrl":"10.1016/j.nucengdes.2024.113742","url":null,"abstract":"<div><div>Preliminary analysis of MgO-BeO composite material used as a moderator in a 50 MWt block-type high-temperature gas-cooled reactor (HTGR) was performed in our previous study. The target burnup of 80 GWd/t was achieved with a uniform fuel composition of 17 wt% ²³⁵U enrichment and 6 kg of heavy metal per fuel block. However, this resulted in high excess reactivity and a peak in axial power distribution at the core center. Therefore, this study aims to reduce excess reactivity by incorporating burnable poison (BP) material and optimize the axial power profile by introducing a nonuniform fuel composition in the core. Neutronic calculations were performed using the Monte Carlo MVP3.0 code developed by the Japan Atomic Energy Agency (JAEA). In this study, three fuel enrichments of ²³⁵U, ranging from 15 wt% to 20 wt%, were distributed across the core while maintaining a constant fuel packing fraction of 45 %. The results showed that the higher power density distribution shifted from the core’s center to its upper part, leading to lower power density in the bottom region than the top. In addition, excess reactivity was reduced by inserting BP rods. Several parametric calculations were performed to achieve minimal excess reactivity without compromising the burnup target. The results showed that the BP rod with a radius of 0.7 cm and 12 wt% of Gd₂O₃ can reduce the excess reactivity from 25.5 %Δk/k to 13.47 % Δk/k.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113742"},"PeriodicalIF":1.9,"publicationDate":"2024-11-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142744118","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 novel approach for full-core mesh generation to enable high-fidelity thermal-hydraulic simulation of nuclear reactor engineering","authors":"Xue Miao,&nbsp;Lingyu Dong,&nbsp;Zhaoshun Wang,&nbsp;Lei Zhang,&nbsp;Jialei Wang,&nbsp;Shihe Wang,&nbsp;Yunhan Zhang,&nbsp;Hongzhen Zhang,&nbsp;Fangxiao Zhang,&nbsp;Changjun Hu","doi":"10.1016/j.nucengdes.2024.113684","DOIUrl":"10.1016/j.nucengdes.2024.113684","url":null,"abstract":"<div><div>Thermal-hydraulic analysis is crucial in reactor engineering. High-fidelity simulations, utilizing advanced computing techniques and supercomputing resources, are highly regarded. High-quality fluid mesh models are essential for complex reactors’ high-fidelity simulations. Using existing tools for model construction has limitations in quality control, performance, user dependency, file generation, and visualization. Estimating time and memory consumption for full-core meshing is also not possible. A R-IMG approach is designed, it effortlessly creates mesh models for intricate flow field, demonstrating exceptional modeling performance, robustness, scalability, and reduced user dependency, while its flexible file manner effectively addresses challenges in generating and visualizing large-scale mesh files. Extensive testing validates R-IMG’s effectiveness and reliability in meshing the reactor’s flow field. It efficiently generates high-quality meshes for the complex flow field in the entire fuel region of CEFR, completing the process within 7 h and 10GB of memory. The resulting model has around 14 billion cells and an average quality of 0.7. R-IMG achieves a maximum parallel scale of 3200 processes for file generation, with approximately 90% parallel efficiency. These results demonstrate that R-IMG outperforms existing tools in core meshing and shows significant potential for full-core meshing. Successful visualization of models and benchmark tests provide evidence for models’ correctness.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113684"},"PeriodicalIF":1.9,"publicationDate":"2024-11-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142744119","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":"Preliminary investigations into optimizing detector configuration for nondestructive assay of fresh nuclear fuel","authors":"Cathleen Barker,&nbsp;Brice Turner ,&nbsp;Andreas Enqvist","doi":"10.1016/j.nucengdes.2024.113748","DOIUrl":"10.1016/j.nucengdes.2024.113748","url":null,"abstract":"<div><div>Like other components of the nuclear fuel cycle, fresh nuclear fuel has potential safeguard and security concerns. Non-destructive assay (NDA) measurements need to be examined and made available to nuclear industry partners to decrease safeguard and oversight concerns for nuclear power plants and facilities. This study analyzed four proposed evaluation criteria: quantity of detectors required, speed and accuracy of detector verification, overall efficiency of neutron and gamma detections, and suitability of observables to advanced machine learning algorithms. Four different detectors have been arranged to be evaluated simultaneously: four Saint-Gobain ³He proportional counters, an Eljen EJ-309 liquid scintillator, a Saint-Gobain / Bicron NaI(Tl) scintillation detector, and a Cs₂LiYCl₆:Ce³⁺(CLYC) scintillation detector. Mock-up BWR and PWR fresh fuel configurations were developed by creating 4x4 and 3x3 natural uranium assemblies and aided in assessments for detector configurations. Additionally, ²⁵²Cf was included to increase the number of neutrons for evaluation. The results of this work will be used to develop a versatile, multi-detector system for fresh fuel verification. The exploration of detector types, data modalities, and their significance as input for advanced machine learning and data analysis algorithms is being explored to understand the obstacles and opportunities in the development of a consolidated low-cost, commercial-off-the-shelf system that nuclear power plants, facilities, and/or inspectors can use for non-destructive verification of fresh nuclear fuel and assembly templating for quick nuclear material inventory verification.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113748"},"PeriodicalIF":1.9,"publicationDate":"2024-11-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142744121","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":"Robust controller design strategies with improved performance for MSBR core","authors":"Akanksha Dwivedi,&nbsp;Ahmad Ali,&nbsp;Chanya","doi":"10.1016/j.nucengdes.2024.113711","DOIUrl":"10.1016/j.nucengdes.2024.113711","url":null,"abstract":"<div><div>Designing controllers for power regulation of nuclear reactors is a challenging task because of its non-linear behavior, and fluctuation of system parameters over time in response to changes in power levels. Motivated by the aforementioned fact, proportional-integral (PI) controllers are designed in this manuscript to regulate the core power of the molten salt breeder reactor (MSBR) by employing the all stabilizing region and quantitative feedback theory (QFT) technique. In the first approach, the entire stability region is plotted in the (<span><math><mrow><msub><mrow><mi>K</mi></mrow><mrow><mi>p</mi></mrow></msub><mo>,</mo><msub><mrow><mi>K</mi></mrow><mrow><mi>i</mi></mrow></msub></mrow></math></span>)-plane. In order to ensure safe operation, the reactivity of control rods must lie in the specified range. Hence, the control effort constrained stability region (CECSR) is obtained within the all stability region to design the PI controller. A novel PI controller and a pre-filter based on gain-phase shaping on Nichols chart are also proposed. Robustness, disturbance rejection and relative stability of the system are addressed by obtaining the performance bounds on Nichols chart. Simulation results illustrate that the closed-loop system is exhibiting a noteworthy improvement in the closed loop performance over the latest reported control strategies.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113711"},"PeriodicalIF":1.9,"publicationDate":"2024-11-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142744068","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":"Inversely estimating flow characteristics of geopolymer paste based on the MPH-I calculations","authors":"Masahiro Kondo ,&nbsp;Sui Satomi ,&nbsp;Ryo Yokoyama ,&nbsp;Shunichi Suzuki","doi":"10.1016/j.nucengdes.2024.113731","DOIUrl":"10.1016/j.nucengdes.2024.113731","url":null,"abstract":"<div><div>A method for inversely predicting the flow properties of Bingham fluid, i.e., yield stress and plastic viscosity of the geopolymer paste, from a simple dam-break experiment was developed based on the MPH-I ((Moving Particle Hydrodynamics for Incompressible flows) calculations. In the decommissioning of Fukushima-Daiichi nuclear power plant, geopolymer is one of the candidate materials for stabilizing the fuel debris and for sealing the PCV (Primary Containment Vessel), where it is important to predict and control the fluidity of the paste of the material. Therefore, the flow property is to be confirmed just before practically pouring the paste. In the inverse estimation, the MPH-I calculations are used only in the preparation phase to obtain the base curve of the front position history with respect to the various yield stresses. Then, both the yield stress and plastic viscosity are estimated using the base curve with an assumption that the inertial force is negligible. Since no additional calculation is needed in the prediction phase, it is applicable for the checking before the pouring. In this study, the methodology was tested against the geopolymer pastes with adding various amount of silica sand for enlarging viscosity. It was confirmed that the front position history obtained from the experiment was well reproduced by the predicted flow properties. This indicates that the Bingham parameters were inversely estimated well.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"431 ","pages":"Article 113731"},"PeriodicalIF":1.9,"publicationDate":"2024-11-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142744120","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}