Nuclear Engineering and Design最新文献_第7页

Development of a transient analysis model for liquid metal cooled space nuclear reactor power system 液态金属冷却空间核反应堆动力系统暂态分析模型的建立

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-30 DOI: 10.1016/j.nucengdes.2025.114089

Li Huaqi, Jiang Xinbiao, Tian Xiaoyan, Zhu Lei, Shi Leitai, Chen Sen, Luo Xiaofei, Li Da, Chen Lixin

{"title":"Development of a transient analysis model for liquid metal cooled space nuclear reactor power system","authors":"Li Huaqi, Jiang Xinbiao, Tian Xiaoyan, Zhu Lei, Shi Leitai, Chen Sen, Luo Xiaofei, Li Da, Chen Lixin","doi":"10.1016/j.nucengdes.2025.114089","DOIUrl":"10.1016/j.nucengdes.2025.114089","url":null,"abstract":"<div><div>The space nuclear reactor power system (SNRPS), which uses a liquid metal-cooled reactor coupled with a closed-loop Brayton cycle for thermoelectric conversion, has emerged as the preferred choice for an advanced space power system. This preference is due to its excellent heat transfer efficiency in the core, wide power range, controllable equipment size, high power conversion efficiency, and mature technology. This study focuses on establishing a system transient analysis model for reactor design, control, and safety analysis of a SNRPS. The model includes subsystem models such as thermal–hydraulic models for the reactor core and primary coolant loop systems, power conversion unit models, and heat pipe radiator models. Six types of benchmark test problems are used to validate each sub-module and component model. The results show that the maximum absolute error between the sub-module model and the analytical solutions of these benchmark test problems is within 2%. Based on the theoretical model established, a transient analysis code for the space nuclear reactor (TACSNR) was developed. The TACSNR was verified using steady-state design parameters from the ultra-small liquid metal cooled space nuclear reactor power system concept (ULCR SNRPS) and the inherent safety sectored compact reactor with a SiGe thermoelectric (TE) power conversion assembly space nuclear reactor power system (SCoRe-TE SNRPS) startup transient process. The calculation results show that the maximum absolute deviation between the calculated values of the TACSNR and the steady-state design parameters of the ULCR SNRPS conceptual scheme is less than 1%, consistent with the parameter change trend and numerical values during the transient startup process of the SCoRe-TE SNRPS system. Additionally, the maximum relative deviation at rated steady state of the SCoRe-TE SNRPS is less than 12%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114089"},"PeriodicalIF":1.9,"publicationDate":"2025-04-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143890521","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

Development of point model for recombiner and its validation with tests conducted in THAI facility using CFD 基于CFD的重组器点模型的建立及其在泰国设备上的验证

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-29 DOI: 10.1016/j.nucengdes.2025.114100

Rupak Kumar Raman , R.S. Rao , Kannan N. Iyer , Sanjeev Gupta

{"title":"Development of point model for recombiner and its validation with tests conducted in THAI facility using CFD","authors":"Rupak Kumar Raman , R.S. Rao , Kannan N. Iyer , Sanjeev Gupta","doi":"10.1016/j.nucengdes.2025.114100","DOIUrl":"10.1016/j.nucengdes.2025.114100","url":null,"abstract":"<div><div>A point model for a recombiner compatible for CFD code to simulate hydrogen distribution and mitigation in nuclear power plant containments is developed in this study. Given the significantly smaller size of the recombiner compared to containment compartments, fully resolving the recombination process requires detailed geometric modeling and a highly refined mesh within the recombiner channels, which require high computational demand. Present study simplifies the recombiner channel into a single computational node while ensuring conservation of momentum, energy and species to address the issue of computational demand. A set of correlations for pressure drop, heat generation and heat transfer coefficients has been developed for recombiner channel in the presence of a steam environment through parametric studies. This model is incorporated in CFD code FLUENT using UDF. Through various simulations the derived point model is verified under both steady-state and transient condition with the detailed computational model. Further validation is performed using CFD simulations of the HR-49, HR 2 and HR 5 tests from the THAI experiment program for various test conditions. The simulation with the point model successfully predicted the recombiner performance and species concentration with experimental data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114100"},"PeriodicalIF":1.9,"publicationDate":"2025-04-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143882779","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

Loss-of-flow transients in the liquid metal-cooled reactor-pool experiment E-SCAPE 液态金属冷却堆池实验中的失流瞬态

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-28 DOI: 10.1016/j.nucengdes.2025.114086

Katrien Van Tichelen , Fabio Mirelli , Yann Bartosiewicz , William D’haeseleer

{"title":"Loss-of-flow transients in the liquid metal-cooled reactor-pool experiment E-SCAPE","authors":"Katrien Van Tichelen , Fabio Mirelli , Yann Bartosiewicz , William D’haeseleer","doi":"10.1016/j.nucengdes.2025.114086","DOIUrl":"10.1016/j.nucengdes.2025.114086","url":null,"abstract":"<div><div>Natural circulation in the primary system is one of the key mechanisms for removing the decay heat in the MYRRHA pool-type research reactor under development at SCK CEN, the Belgian Nuclear Research Centre. To confirm the feasibility of this passive approach, experiments are performed in the E-SCAPE facility, a thermal hydraulic 1/6-scale 3-D model of the primary system of MYRRHA, with an electrical core simulator and cooled with Lead Bismuth Eutectic.</div><div>This paper presents the outcome of transient loss-of-flow (LOF) experiments in E-SCAPE. First, the representativeness of LOF transients in E-SCAPE for MYRRHA is demonstrated based on simplified analytical integral models of the reactor prototype and of the scaled facility. Next, results of several test cases with varying core powers and system pressure losses are reported. In all cases studied, a smooth transition from forced to buoyancy-driven natural circulation is observed after the LOF event, with the establishment of stable, lower flow rates. Decay heat can be safely removed from the core as the maximum core temperatures stay within safety limits. Two phases can be identified during the transient: an initial phase dominated by mass inertia, and a second phase dominated by the heat capacity of the system. The final steady state shows significant thermal stratification in the upper plenum.</div><div>The extensive instrumentation in the E-SCAPE facility allows direct comparison of experimental data with numerical simulations, allowing validation of simulation tools in representative conditions. This is essential for the safety assessment and licensing process of MYRRHA.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114086"},"PeriodicalIF":1.9,"publicationDate":"2025-04-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143882100","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

Simulating experimentally observed nonlinear response of large-scale concrete structure to understand the selection of damping: A case of minor nonlinearities 模拟实验观测的大型混凝土结构的非线性响应以理解阻尼的选择：一个轻微非线性的例子

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-26 DOI: 10.1016/j.nucengdes.2025.114098

Sangwoo Lee, Abhinav Gupta, Giorgio T. Proestos

{"title":"Simulating experimentally observed nonlinear response of large-scale concrete structure to understand the selection of damping: A case of minor nonlinearities","authors":"Sangwoo Lee, Abhinav Gupta, Giorgio T. Proestos","doi":"10.1016/j.nucengdes.2025.114098","DOIUrl":"10.1016/j.nucengdes.2025.114098","url":null,"abstract":"<div><div>Recent studies conducted by the US Nuclear Regulatory Commission and its collaborators have explored the use of limit-state C for SDCs 5 and 4, unlike the conventional design of concrete shear walls in nuclear power plants. Consideration of the limit-state C allows minor nonlinearity in the behavior of structural systems when subjected to design earthquakes. In the context of the nonlinear behavior in concrete structures, the selection of appropriate parameters for the concrete’s constitutive material model is important. In addition, there are some concerns with using Rayleigh damping in nonlinear seismic analysis because, many studies have shown that an improper use of Rayleigh damping in the nonlinear seismic analysis can lead to unintended large damping forces thereby resulting in an underestimation of response parameters. In this study, response data from a large-scale shake table experiment of a 3-story concrete shear wall structure is used to understand these effects. A finite element analysis of the test specimen using concrete damage plasticity model and its reconciliation with the experimental data is used to understand two aspects discussed above, i.e., (i) selection of model parameters in the Concrete Damage Plasticity Model for nonlinear seismic analysis of concrete structures, and (ii) selection of an appropriate damping model. Both of these aspects are studied for the case of minor damage (nonlinearity) in the structure corresponding to ASCE-43′s guidelines for risk-informed performance-based design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114098"},"PeriodicalIF":1.9,"publicationDate":"2025-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143877038","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

Design and pre-test analyses of an integral thermal–hydraulic facility for a prismatic gas-cooled micro reactor 柱形气冷微堆整体热水力装置设计与预试分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-26 DOI: 10.1016/j.nucengdes.2025.114097

Zheng Huang , Miaoxin Jiao

{"title":"Design and pre-test analyses of an integral thermal–hydraulic facility for a prismatic gas-cooled micro reactor","authors":"Zheng Huang , Miaoxin Jiao","doi":"10.1016/j.nucengdes.2025.114097","DOIUrl":"10.1016/j.nucengdes.2025.114097","url":null,"abstract":"<div><div>To support the R&D of the novel gas-cooled micro reactor (GMR), an integral thermal–hydraulic test facility (named INTALI) is designed to explore the key thermal–hydraulic phenomena specific to the GMR, thereby providing necessary data to validate computer codes for thermal–hydraulic and accident transient analysis. This paper presents a preliminary design of the INTALI facility, experimental methodology, and pre-test analyses. The INTALI facility consists of a primary loop operating at the prototypical temperature and pressure and a test section containing a scaled-down simulated reactor and a passive core cooling system (PCCS). Steady-state and transient tests will be carried out, which correspond to the normal operation and the pressurized loss of forced coolant (PLOFC) accident condition of the GMR, respectively. The experiment is mainly to investigate: (i) the coupling between the reactor and the PCCS, especially during the PLOFC, (ii) the operational characteristics of the PCCS and the energy distribution, and (iii) potential thermal stratification in the gravitational direction caused by the natural circulation in the PCCS. The pre-test analyses of the experiment were performed by CFD simulations using the COMSOL Multiphysics software. The predicted 3D distributions of the temperature and velocity fields for both the reactor and the PCCS are used to determine the instrumentation scheme. The simulation results show that no significant vertical thermal gradient is observed on the RPV wall. The radiative heat transfer from the RPV to the PCCS insulation layer plays an important role in heat removal in addition to convection. The heat removal capability of the PCCS is significantly influenced by the RPV’s temperature during the PLOFC transient. The developed CFD model is also ready for post-test quantification and validation once the experimental data is available.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114097"},"PeriodicalIF":1.9,"publicationDate":"2025-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143874708","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 hydrodynamic loads on the planar trash-blocking nets at coastal nuclear power plant intake under flow action based on finite element analysis 基于有限元分析的沿海核电站取水口平面拦污栅在水流作用下的水动力负荷评估

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-25 DOI: 10.1016/j.nucengdes.2025.114096

Shaojian Guo , Haoran Liu , Cheng Zhou , Rong Wan , Yucheng Wang , Zhiqiang Liu

{"title":"Evaluation of hydrodynamic loads on the planar trash-blocking nets at coastal nuclear power plant intake under flow action based on finite element analysis","authors":"Shaojian Guo , Haoran Liu , Cheng Zhou , Rong Wan , Yucheng Wang , Zhiqiang Liu","doi":"10.1016/j.nucengdes.2025.114096","DOIUrl":"10.1016/j.nucengdes.2025.114096","url":null,"abstract":"<div><div>The trash-blocking net facility installed at the water intake serves as a crucial barrier for coastal nuclear power plants, safeguarding against the risks posed by marine biofouling. However, the complex marine environment, along with heavy biofouling, poses significant challenges to the safe operation of trash-blocking nets. In this study, a comprehensive investigation of trash-blocking nets under uniform current conditions was conducted using finite element-based numerical simulations, incorporating variations in flow velocity, water level, net width, and solidity ratio. The results indicated that as flow velocity increased, water level decreased, net wider and solidity ratio rose, the total drag, deformation, and tensions in various components (main rope, reinforcing ropes, anchor rings, and twines) exhibited an increasing trend. Based on these findings, empirical expressions were developed to represent the individual and combined effects of the influencing factors on total drag force and main rope tension. The total drag force and main rope tension were found to have a quadratic relationship with flow velocity, water level and solidity ratio, and a linear relationship with net width. This study provides a foundational reference and data support for predicting the loads on trash-blocking nets and preventing failure risks.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114096"},"PeriodicalIF":1.9,"publicationDate":"2025-04-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143869127","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

Comparative evaluation and selection of heat exchangers using multicriteria decision-making 多准则决策下换热器的比较评价与选择

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-25 DOI: 10.1016/j.nucengdes.2025.114066

JunSoo Yoo , Sunming Qin , Silvino A. Balderrama Prieto , Erik Hisahara , Hansol Kim

{"title":"Comparative evaluation and selection of heat exchangers using multicriteria decision-making","authors":"JunSoo Yoo , Sunming Qin , Silvino A. Balderrama Prieto , Erik Hisahara , Hansol Kim","doi":"10.1016/j.nucengdes.2025.114066","DOIUrl":"10.1016/j.nucengdes.2025.114066","url":null,"abstract":"<div><div>This study presents a well-structured method for comparing and selecting HE technologies for an IES. The decision to select a HE for a particular IES configuration can vary greatly depending not only on engineering requirements but also on the customer’s specific demand. In other words, the HE selection for IES requires a multicriteria decision- making approach, taking into account diverse technical, economic, and safety aspects, as well as the relative priorities considered by energy users. This study employs a HE evaluation approach combining multicriteria decision-making techniques widely used in various industries: QFD and AHP techniques. Of particular interest is the use of the proposed method to select a high-temperature HEs that couples advanced nuclear reactors and industrial processes.</div><div>To build a practical basis for comparing HEs within the proposed framework, efforts were made to identify the various HEs requirements for IES purposes. In addition, leveraging the insights obtained from the literature review and the market survey of commercial HE suppliers, a knowledge base was built to facilitate the comparison of each requirement across various HE designs. Also, evaluation metrics were identified for HE requirements with a robust rationale to enhance the quality of decisions made throughout the proposed evaluation process. The evaluation procedure and knowledge base described in this study can provide a useful basis for those interested in screening the appropriate HE designs for various IES scenarios.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114066"},"PeriodicalIF":1.9,"publicationDate":"2025-04-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143874706","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

Flow oscillation fade-out and pool water level effect experiments on open loop passive cooling system 开环被动冷却系统流动振荡淡出及池水位效应试验

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-25 DOI: 10.1016/j.nucengdes.2025.114087

Joonas Telkkä, Vesa Riikonen, Antti Räsänen, Eetu Kotro, Juhani Hyvärinen

{"title":"Flow oscillation fade-out and pool water level effect experiments on open loop passive cooling system","authors":"Joonas Telkkä, Vesa Riikonen, Antti Räsänen, Eetu Kotro, Juhani Hyvärinen","doi":"10.1016/j.nucengdes.2025.114087","DOIUrl":"10.1016/j.nucengdes.2025.114087","url":null,"abstract":"<div><div>The stable operating conditions for an open loop passive containment heat removal system were identified through testing conducted with the PASI test facility, a half-height wall condenser model at LUT University, Finland. Previous tests have shown that open loop systems tend to operate in a quasi-steady oscillatory mode characterized by geysering and flashing. The cessation of flow oscillations depends on the sparger structure. When flooding of the riser pipeline is prevented, the oscillation fade-out and steady two-phase natural circulation is reached quickly after the system reaches saturation conditions. Conversely, if flooding is allowed, the oscillations disappear only at heating power large enough to meet the countercurrent flow limitation (CCFL) criterion in the riser. The impact of gravity head on the system behavior was also examined. The amplitude of two-phase flow oscillations decreased along the lowering of the pool water level. When the water level decreased below the pressure balancing hole, the flow behavior changed since the riser flooding ended. Additionally, the riser boil-out was tested. The results show that the open-loop natural circulation system can effectively remove heat as long as there is water inventory inside the loop, even if the pool is empty of water. The containment pressure rises only when boiling initiates in the heat exchanger.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114087"},"PeriodicalIF":1.9,"publicationDate":"2025-04-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143869128","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

Numerical simulation on thermal-hydraulic-mechanical coupling of core corium migration during severe accidents 重大事故堆芯迁移热-水-力耦合数值模拟

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-25 DOI: 10.1016/j.nucengdes.2025.114083

Yujian Huang, Zhengyang Dong, Mingjun Wang, Kui Zhang, Suizheng Qiu, G.H. Su, Wenxi Tian

{"title":"Numerical simulation on thermal-hydraulic-mechanical coupling of core corium migration during severe accidents","authors":"Yujian Huang, Zhengyang Dong, Mingjun Wang, Kui Zhang, Suizheng Qiu, G.H. Su, Wenxi Tian","doi":"10.1016/j.nucengdes.2025.114083","DOIUrl":"10.1016/j.nucengdes.2025.114083","url":null,"abstract":"<div><div>Core corium migration is one of the critical challenges of severe core accidents. The lower support plate is a critical load-bearing component within the pressure vessel among core corium migration. Investigating the thermal exchange and mechanical failure of the lower support plate during the core migration process holds significant practical value. The heat transfer between the corium and the support plate is complex, involving multiple phenomena such as fluid dynamics, thermal exchange, melting, and mechanical effects, making a comprehensive analysis of the failure process challenging. In this study, a migration heat transfer model has been established, incorporating radiation heat transfer, impact heat transfer, and direct contact heat transfer. The interaction between the corium and the support plate is modeled using a mechanical analysis approach, while the mechanical effects are analyzed through the formulation of a constitutive equation. A Thermal-Hydraulic-Mechanical (THM) coupling calculation method is also developed to address these interactions. The results show that the corium migration heat transfer is consistent with findings in the relevant literature. The majority of corium migrates close to the wall of the RPV lower head, causing the temperature at the edges of the lower support plate to exceed that at the center, leading to creep failure under thermal stress. As the corium continues to migrate, the cumulative mass of molten material and the convective heat transfer coefficient increase. At 60 s, the maximum total deformation of the support plate reaches 0.89025 mm, with a maximum total strain of 0.01636 mm/mm. The equivalent stress is concentrated at the upper surface edges, exceeding the yield limit, indicating fracture failure. Ultimately, the support plate fails within 1 min due to sustained radiation heat. These simulation results offer insights for the safe design of the lower head.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114083"},"PeriodicalIF":1.9,"publicationDate":"2025-04-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143869165","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

Validation of a system code (GAMMA+) using standard k-ε model for multi-dimensional turbulent flows in various geometries 使用标准k-ε模型对不同几何形状的多维湍流进行系统代码（GAMMA+）验证

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-04-25 DOI: 10.1016/j.nucengdes.2025.114079

Seung Hyun Yoon, Nam-il Tak, Hong Sik Lim

{"title":"Validation of a system code (GAMMA+) using standard k-ε model for multi-dimensional turbulent flows in various geometries","authors":"Seung Hyun Yoon, Nam-il Tak, Hong Sik Lim","doi":"10.1016/j.nucengdes.2025.114079","DOIUrl":"10.1016/j.nucengdes.2025.114079","url":null,"abstract":"<div><div>The GAMMA+ (General Analyzer for Multi-component and Multi-dimensional Transient Application) has been developed as a safety analysis tool for non-light water reactors (non-LWRs). Multi-dimensional turbulence modeling capabilities in system codes are essential for analyzing rapid transients in non-LWR nuclear systems with large cores operating in turbulent regimes. While some system codes employ the mixing-length model due to its implementation simplicity, the standard k-ε model is preferred for its superior accuracy and robustness in practical flow applications. This study presents the implementation of the standard k-ε model into GAMMA+, utilizing a square matrix that consists of the temporal differences of the pressure, the temperature, k and ε. Validation of the implemented model encompassed various single-phase flow configurations: flows in a pipe, a plate channel, and a backward-facing step with adiabatic wall conditions; forced convection flows including a pipe, an abruptly expanded pipe, and a backward-facing step with heat flux conditions; and natural convection flows in cavities with fixed temperature boundaries. Comparative analyses against experimental data, Direct Numerical Simulation (DNS) results, and Reynolds-Averaged Navier Stokes (RANS) simulations from well-known computational fluid dynamics (CFD) codes demonstrate the successful implementation of the standard k-ε model in GAMMA+ for multi-dimensional turbulent flow simulations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114079"},"PeriodicalIF":1.9,"publicationDate":"2025-04-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143869126","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