Nuclear Engineering and Design最新文献_第6页

Parametric reduced-order modeling of once-through steam generator via double proper orthogonal decomposition 通过双正交分解对一次通过式蒸汽发生器进行参数化减阶建模

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-07 DOI: 10.1016/j.nucengdes.2024.113627

Yifan Xu, Minjun Peng, Genglei Xia, Xiaobo Zeng

{"title":"Parametric reduced-order modeling of once-through steam generator via double proper orthogonal decomposition","authors":"Yifan Xu, Minjun Peng, Genglei Xia, Xiaobo Zeng","doi":"10.1016/j.nucengdes.2024.113627","DOIUrl":"10.1016/j.nucengdes.2024.113627","url":null,"abstract":"<div><div>Mastering thermal–hydraulic characteristics of the once-through steam generator (OTSG) is essential for ensuring the stable operation and safety of reactors. While refined simulation models offer relatively accurate predictions for OTSG thermal–hydraulic research, the high computational cost often limits their applicability in system online- monitoring and real-time control. Specifically, the computational burden of these models can be prohibitive for multi-query simulation tasks such as optimization design and uncertainty analysis. Model order reduction (MOR) provides a solution that meets the need for both precision and speed in nuclear reactor system. Proper orthogonal decomposition (POD), as one of the representative MOR methods, has been widely used in reactor-related research, but the data-driven reduced order model (ROM) shows poor robustness when applied to situations that deviate from the modeling conditions. Therefore, a parametric ROM suitable for estimating the thermal and hydraulic characteristics of OTSG is established in this work by introducing double POD (DPOD). The model is verified based on the full-order model (FOM) developed in the RELAP5 code. Verification results demonstrate that the maximum relative error between the ROM estimations and FOM data is less than 0.5%, while the computational time of the ROM is less than 0.1 s. This parametric ROM thus satisfies the requirements for efficient and accurate estimation of OTSG thermal–hydraulic characteristics, providing a viable alternative to refined simulation models for multi-query simulation tasks and supporting for nuclear digital twins.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142428547","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}

引用次数: 0

Scaling design and similarity analysis of a floating nuclear power plant 浮动核电站的规模设计和相似性分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-07 DOI: 10.1016/j.nucengdes.2024.113624

Qingsong Liu , Yuexin Wang , Xi Deng , Chen Xi , Cheng He , Daoyi Chen , Huan He

{"title":"Scaling design and similarity analysis of a floating nuclear power plant","authors":"Qingsong Liu , Yuexin Wang , Xi Deng , Chen Xi , Cheng He , Daoyi Chen , Huan He","doi":"10.1016/j.nucengdes.2024.113624","DOIUrl":"10.1016/j.nucengdes.2024.113624","url":null,"abstract":"<div><div>In the process of developing floating reactor systems, their operation in the marine environment needs to be simulated to ensure the reliability of the safe design of nuclear reactors. Scaling test is a potential option for computer model validation by virtue of its low cost and flexibility. This study first gives the similarity law for dynamic tests of the reactor system in high-temperature environments by dimensional analysis. Then we focus on analyzing the jamming phenomenon that may occur during the similarity design process and provides a solution for realizing the contact state of the prototype in the scaled model. This study also analyzes the dynamic response of a floating reactor system under extreme operating conditions by combining numerical simulations and scaling tests. By comparing the response results predicted by the scaled model with the prototype response, it is further demonstrated that the feasibility of scaling test as an alternative to full-size test to provide a reference for realizing low-cost, accurate and safe design for reactor system.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142428801","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}

引用次数: 0

Monitoring outer surface temperature to assess thermal fatigue at T-junctions 监测外表面温度以评估 T 型接头的热疲劳

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-05 DOI: 10.1016/j.nucengdes.2024.113622

Koji Miyoshi

引用次数: 0

3D topology optimization design of air natural convection heat transfer fins 空气自然对流传热翅片的三维拓扑优化设计

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-05 DOI: 10.1016/j.nucengdes.2024.113623

ChuanChang Dong , ChunBo Zhang , GeNing He , DongHui Li , ZiWei Zhang , JiDong Cong , ZhaoMing Meng , Shehzad Asim , Mehtab Ashraf

{"title":"3D topology optimization design of air natural convection heat transfer fins","authors":"ChuanChang Dong , ChunBo Zhang , GeNing He , DongHui Li , ZiWei Zhang , JiDong Cong , ZhaoMing Meng , Shehzad Asim , Mehtab Ashraf","doi":"10.1016/j.nucengdes.2024.113623","DOIUrl":"10.1016/j.nucengdes.2024.113623","url":null,"abstract":"<div><div>This study focuses on the fin-tube heat exchanger and utilizes topology optimization methods to design a completely new fin structure. In this optimization process, complete Navier-Stokes (N-S) equations were used to describe the steady-state incompressible flow, and the Boussinesq model was employed to simulate natural convection. The flow equations were coupled with the heat convection–diffusion equation to achieve topology optimization for natural convection heat transfer. Topology optimization was conducted using density-based optimization methods, and interpolation was performed on the permeability and conductivity of the distributed materials. Given the initial fin structure, an interpolation progressive approach was adopted to obtain a new “ airfoil-shaped” optimized structure through density-based topology optimization method for natural convection. The new structure enhances the convective heat transfer by perforating the fins. The perforations are mainly concentrated in the central region of the heat exchanger and the upper half of the fins. The new structure, compared to the prototype structure, not only has a reduced volume but also exhibits a decrease in convective thermal resistance within a larger range of heat flux densities, as revealed by CFD simulations. Moreover, as the heat flux density increases, the rate of reduction in convective thermal resistance shows an upward trend for the new structure compared to the prototype structure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142428805","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}

引用次数: 0

Application of release starting time classification for planning emergency preparedness and response to the hypothetical accident scenario of iPWR-SMR in Thailand 在泰国 iPWR-SMR 假设事故情景中应用释放起始时间分级来制定应急准备和响应计划

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-05 DOI: 10.1016/j.nucengdes.2024.113629

Wasin Vechgama , Piyawan Krisanangkura , Kampanart Silva

{"title":"Application of release starting time classification for planning emergency preparedness and response to the hypothetical accident scenario of iPWR-SMR in Thailand","authors":"Wasin Vechgama , Piyawan Krisanangkura , Kampanart Silva","doi":"10.1016/j.nucengdes.2024.113629","DOIUrl":"10.1016/j.nucengdes.2024.113629","url":null,"abstract":"<div><div>Due to the interest in SMR reactors in newcomer countries, the understanding of the risk of source term release and dose exposure of SMR technology is important scientific data for communicating between the government and people. This study aims to extend the application of release starting time classification of level 2 PSA in SMR technology to inform strategic planning for nuclear consequences and determine size requirements for the emergency planning zones in level 3 PSA. The SBO accident scenario of iPWR at the location in the Ubon Ratchathani province, Thailand, was investigated in this study. The GMM is used to classify the probability density of uneven distributions of release starting times into the two groups. The higher probability density and maximum radioactive release in Group (1) were used to suggest the main plan for emergency response. In the main plan, the local government needs to evacuate the people outside 6 km to avoid dose exposure if source term release is monitored within 8–21 h. The impact of source term release in Group (2) was set as a backup plan for considering an emergency planning extension if the time delay to later than 21 h. Finally, the nuclear consequences of SMRs are compared with large NPPs in the same accident scenarios. SMR technology has the potential to support flexible emergency planning zones for sheltering and evacuation without significant dose exposure to neighboring countries.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142428806","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}

引用次数: 0

Performance of a hydraulic buffer for PWR fuel assemblies: Mathematical modeling, numerical solutions, and experimental comparison 压水堆燃料组件液压缓冲器的性能：数学建模、数值求解和实验对比

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-04 DOI: 10.1016/j.nucengdes.2024.113619

Heng Huang, Peng Li, Chenguang Fan

{"title":"Performance of a hydraulic buffer for PWR fuel assemblies: Mathematical modeling, numerical solutions, and experimental comparison","authors":"Heng Huang, Peng Li, Chenguang Fan","doi":"10.1016/j.nucengdes.2024.113619","DOIUrl":"10.1016/j.nucengdes.2024.113619","url":null,"abstract":"<div><div>In a Pressurized Water Reactor (PWR), the hydraulic buffer serves as an essential component, significantly mitigating the impact force between the control rod drive mechanism and the fuel assembly in scenarios of emergency shutdown. This paper provides a complete analysis of the dynamical performance of a new type of hydraulic buffer, including its mathematical modeling, numerical solution scheme, and experimental comparison. During the fluid modeling, five typical cases are first classified in terms of both flow directions and coefficients based on the relative positions of the sleeve and the piston. A new flow iterative calculation format encompassing flow directions that can efficiently solve the flow coefficients is proposed during the fluid modeling. A one-way coupling scheme is used for the fluid–structure dynamical solutions. An experimental comparative study is conducted using the standard and reference (SR) case, and the present method shows good agreement with the experiment. In the present analysis: (1) The dynamic characteristics of the buffer are fully demonstrated using time history curves and phase diagrams, the physical parameters of fluid and structural motions, mainly including fluid pressure inside the chamber, dynamic relative displacement and velocity of the piston and sleeve, the rebound disparagement, and system kinetic energy; (2) Typical features of double peaks of the impact force, related to the collision between the piston, the impacted object, and the sleeve, have been captured and well simulated; (3) The variation of the two peaks of the impact forces with parameters and the transformation laws of the maximum impact force between these two peaks have been fully revealed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142428803","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}

引用次数: 0

Mechanistic prediction of Westinghouse TRITON11®# BWR fuel critical power with MEFISTO-T subchannel analysis code 利用 MEFISTO-T 子信道分析代码对西屋 TRITON11®# 核燃料堆燃料临界功率进行机理预测

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-04 DOI: 10.1016/j.nucengdes.2024.113613

Jean-Marie Le Corre

{"title":"Mechanistic prediction of Westinghouse TRITON11®# BWR fuel critical power with MEFISTO-T subchannel analysis code","authors":"Jean-Marie Le Corre","doi":"10.1016/j.nucengdes.2024.113613","DOIUrl":"10.1016/j.nucengdes.2024.113613","url":null,"abstract":"<div><div>The <strong>TRITON11®</strong> fuel design is the latest Boiling Water Reactor (BWR) fuel product developed by Westinghouse, based on an 11 × 11 optimized fuel rod lattice, including mixing vane spacer grids, three large water rods and 18 part-length rods. The design offers larger fuel cycle cost saving, improved fuel reliability and increased thermal margin over previous Westinghouse fuel products. The critical power performances of the TRITON11 fuel design were assessed at the Westinghouse thermal–hydraulic FRIGG loop using a full-scale test bundle covering a wide range of BWR core conditions (covering normal operation and Anticipated Operational Occurrences) with various radial and two axial power distributions. The resulting FRIGG steady state database was simulated with Westinghouse subchannel analysis code MEFISTO-T based on a two-phase three-field approach of annular two-phase flow, accounting for the drop deposition enhancement provided by the spacer grids. A new model of local film entrainment was introduced due to the liquid “scrapping off” effects provided by the frame and side vanes of the spacer grids along the fuel channel and water rods. After spacer grid effect calibration, the steady state critical power is simulated mechanistically by power iterations up to complete local film dryout. The MEFISTO-T code can successfully predict the critical power performance of the Westinghouse TRITON11 BWR fuel design, including on part-length rods, with almost no bias and trend, within a standard deviation of about 5 %.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142428804","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}

引用次数: 0

Experimental and numerical studies on the hydraulic performance and the internal flow characteristics of lead-bismuth in the main coolant pump of LFR 对低温冷冻机主冷却剂泵中铅铋的水力性能和内部流动特性的实验和数值研究

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-01 DOI: 10.1016/j.nucengdes.2024.113618

Zutao Xiang, Liangxing Li, Xiangyang Xu, Shang Shi, Zhenxin Lei

{"title":"Experimental and numerical studies on the hydraulic performance and the internal flow characteristics of lead-bismuth in the main coolant pump of LFR","authors":"Zutao Xiang, Liangxing Li, Xiangyang Xu, Shang Shi, Zhenxin Lei","doi":"10.1016/j.nucengdes.2024.113618","DOIUrl":"10.1016/j.nucengdes.2024.113618","url":null,"abstract":"<div><div>As the key component of the primary circuit of a lead-cooled fast reactor (LFR), the hydraulic performance and the internal flow characteristics of the main coolant pump (MCP) are significantly important to the long-term safe and efficient operation of LFR. The present study reports an experimental and numerical investigation of the hydraulic performance and the internal flow characteristics of a centrifugal pump, which is considered one of the candidates for the MCP in LFR. Based on the experimental platform of Liquid Metal-Supercritical Fluid coupled heat transfer (LMSF) of Xi’an Jiaotong University, the hydraulic performances of a centrifugal pump like the pump head and its efficiency are studied and the measured data is also used to validate the accuracy of the numerical method which are employed to study the flow characteristics of MCP, as velocity, pressure, turbulent kinetic energy (TKE). The results show that the deviation of experimental and numerical simulation results is less than 5%. The high velocity region in the impeller channel is concentrated near the hub side of the blade inlet, and the significant pressure difference of the blade surface is primarily concentrated in the second half and shroud region. The inlet area of the blade exhibits a concentration of high TKE and the circular distribution of TKE within the impeller exhibits a periodic pattern, with its frequency being consistent with the number of blades.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142359536","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}

引用次数: 0

Evaluation of proliferation resistance of SMRs with COMPRE methodology assisted by Pu production 在钚生产的辅助下，利用 COMPRE 方法评估 SMR 的抗增殖能力

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-01 DOI: 10.1016/j.nucengdes.2024.113615

Kwangho Ju , SeoungHo Jeong , Hosik Yoo

{"title":"Evaluation of proliferation resistance of SMRs with COMPRE methodology assisted by Pu production","authors":"Kwangho Ju , SeoungHo Jeong , Hosik Yoo","doi":"10.1016/j.nucengdes.2024.113615","DOIUrl":"10.1016/j.nucengdes.2024.113615","url":null,"abstract":"<div><div>The proliferation resistance (PR) of several small modular reactors (SMR) is investigated in terms of material characteristic and other extrinsic parameters with COMPRE methodology developed by KINAC. In this work, plutonium (Pu) production amount and isotopic yield are newly introduced as key metrics in material characteristic. Their functionalization is performed with a linear regression method based on sparse core depletion data. Nevertheless, a clear discrepancy is observed in Pu buildup and its quality depending on SMR type and core burnup status. Some of the non-light water SMRs exhibit a higher proliferation concern and their PR score trends are clearly different compared to conventional pressurized water reactors (PWR). Other extrinsic parameters are evaluated based on development status of representative SMRs. At present, liquid metal cooled-fast reactor (LMFR) and molten salt reactor (MSR) receive low scores due to their limited technical maturity in reactor design and related I&C equipment. We conclude that distinct safeguards approaches, suitable for a wide range of SMR designs should be developed for commercial use in the near future.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142359539","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}

引用次数: 0

Effect of diffusion barriers on steam oxidation properties and interface evolutions of Cr coating for zirconium alloy at 1200 °C 扩散屏障对锆合金在 1200 °C 下的蒸汽氧化特性和铬涂层界面演变的影响

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-10-01 DOI: 10.1016/j.nucengdes.2024.113614

Xiaodong He , Yu Jin , Zijin Liu , Xiaoze Yue , Guangan Zhang , Shunhua Wang , Lunlin Shang

{"title":"Effect of diffusion barriers on steam oxidation properties and interface evolutions of Cr coating for zirconium alloy at 1200 °C","authors":"Xiaodong He , Yu Jin , Zijin Liu , Xiaoze Yue , Guangan Zhang , Shunhua Wang , Lunlin Shang","doi":"10.1016/j.nucengdes.2024.113614","DOIUrl":"10.1016/j.nucengdes.2024.113614","url":null,"abstract":"<div><div>Under the ATF design concept, Cr-based coating is the most promising zirconium alloy cladding protective coating in the case of PWR accidents. However, the primary factor affecting the oxidation resistance of Cr coatings is the interdiffusion of Cr and Zr. In this study, Cr coatings with three refractory metal interlayers of Nb, Mo, and Ta were deposited on zirconium alloy surfaces using closed-field unbalanced magnetron sputtering. The study investigated the effects of different interlayers on the microstructure, oxidation properties, and interface evolution process of Cr coatings under steam conditions at 1200 ℃, and compared them with Cr coatings without intermediate layers. The results indicate that the refractory metal interlayer influences the orientation and growth rate of Cr coating grains. The formation of a Laves phase mixed layer in the interlayer during oxidation effectively hinders the diffusion of O and Cr to the zirconium alloy matrix, enhancing its oxidation resistance. Furthermore, the diffusion rate and vacancy concentration of Mo and Ta elements to the Zr layer exceed that of diffusion to the Cr layer. During the cooling process, a precipitate phase forms in the Zr-4 matrix, while the Nb coating forms an infinite solid solution in the β-Zr.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142359538","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}

引用次数: 0