Science and Technology of Nuclear Installations最新文献_第10页

An Efficient Scheme for Coupling OpenMC and FLUENT with Adaptive Load Balancing 基于自适应负载均衡的OpenMC与FLUENT的有效耦合方案

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-09-24 DOI: 10.1155/2021/5549602

Qingyang Zhang, Tianji Peng, Guangchun Zhang, Jie Liu, Xiaowei Guo, Chunye Gong, Bo Yang, Xukai Fan

{"title":"An Efficient Scheme for Coupling OpenMC and FLUENT with Adaptive Load Balancing","authors":"Qingyang Zhang, Tianji Peng, Guangchun Zhang, Jie Liu, Xiaowei Guo, Chunye Gong, Bo Yang, Xukai Fan","doi":"10.1155/2021/5549602","DOIUrl":"https://doi.org/10.1155/2021/5549602","url":null,"abstract":"This paper develops a multi-physics interface code MC-FLUENT to couple the Monte Carlo code OpenMC with the commercial computational fluid dynamics code ANSYS FLUENT. The implementations and parallel performances of block Gauss–Seidel-type and block Jacobi-type Picard iterative algorithms have been investigated. In addition, this paper introduces two adaptive load-balancing algorithms into the neutronics and thermal-hydraulics coupled simulation to reduce the time cost of computation. Considering that the different scalability of OpenMC and FLUENT limits the performance of block Gauss–Seidel algorithm, an adaptive load-balancing algorithm that can increase the number of nodes dynamically is proposed to improve its efficiency. Moreover, with the natural parallelism of block Jacobi algorithm, another adaptive load-balancing algorithm is proposed to improve its performance. A 3 x 3 PWR fuel pin model and a 1000 MWt ABR metallic benchmark core were used to compare the performances of the two algorithms and verify the effectiveness of the two adaptive load-balancing algorithms. The results show that the adaptive load-balancing algorithms proposed in this paper can greatly improve the computing efficiency of block Jacobi algorithm and improve the performance of block Gauss–Seidel algorithm when the number of nodes is large. In addition, the adaptive load-balancing algorithms are especially effective when a case demands different computational power of OpenMC and FLUENT.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"48446795","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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: 7

Failure Fraction Calculation of the TRISO-Coated Particle Using X-Ray Computed Tomography 用x射线计算机断层扫描计算triso包覆颗粒的失效分数

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-09-21 DOI: 10.1155/2021/2420526

Libing Zhu, Jianxun Zhao, Xincheng Xiang, Yu Zhou, Xiangang Wang

引用次数: 1

Detection of Illegal Movement in Radioactive Material Transportation Security Systems 放射性物质运输安全系统中非法移动的检测

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-09-09 DOI: 10.1155/2021/6670114

Tiejun Zeng, Xiaohua Yang, Yaping Wan, Panpan Jiang, L. Zhenghai

引用次数: 0

Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method 基于顺序Abaqus-FRANC3D仿真方法的反应堆压力容器完整性评估

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-09-06 DOI: 10.1155/2021/7035787

M. Annor-Nyarko, Hong Xia

{"title":"Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method","authors":"M. Annor-Nyarko, Hong Xia","doi":"10.1155/2021/7035787","DOIUrl":"https://doi.org/10.1155/2021/7035787","url":null,"abstract":"The safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences—inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The ASME fracture toughness of the reactor vessel’s steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"49555313","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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: 3

The Implementation of Diverse Actuation System in ACPR1000 Nuclear Power Plants ACPR1000型核电厂多种驱动系统的实现

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-09-06 DOI: 10.1155/2021/5529570

Zhen-Ying Wang, Zhiyun Liu, T. Ma, Chen Sun, Liu Liu, Yu Huang, Gang-Miao Guo

{"title":"The Implementation of Diverse Actuation System in ACPR1000 Nuclear Power Plants","authors":"Zhen-Ying Wang, Zhiyun Liu, T. Ma, Chen Sun, Liu Liu, Yu Huang, Gang-Miao Guo","doi":"10.1155/2021/5529570","DOIUrl":"https://doi.org/10.1155/2021/5529570","url":null,"abstract":"In order to improve the capability of dealing with software common cause failure (CCF) of digital reactor protection and monitoring system (RPMS), the diverse actuation system (DAS) is introduced for ACPR1000 nuclear power plants. From economic and feasibility point of view, the solution of DAS sharing with RPMS sensors and actuators is suggested; after capturing the function requirement of DAS, the automatic functions and manual functions assigned to it are determined based on transient analysis of design basic accidents concurrent with software CCF of RPMS. The independent verification proves that the reactor can be fallen back to and maintained at safety shutdown state, thanks to these DAS functions. Insight into probabilistic safety assessment proves significant reductions of risks are contributed. The critical technical issues while implementing DAS, such as measures to ensure its diversity from RPMS, precautions for preventing from its spurious actuation, isolation and independency from RPMS, and its testability and maintainability, are deliberately settled to improve its engineering reliability and alleviate the impact on RPMS as far as possible. Field programmable gate array technology that is diversified from RPMS is chosen to build DAS of ACPR1000 nuclear power plant, and the commissioning test verifies that it is capable of performing its designed functions. At last, a set of DAS-specific, paper-based, and event-oriented emergency operating procedure is developed, verified, and validated. Until now, the DAS system has always been successfully operating in all ACPR1000 nuclear power plants for several years.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"46297297","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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: 1

Study on Airborne Radionuclide Dispersion in Floating Nuclear Power Plant under the Loss-of-Coolant Accident 冷却剂丢失事故下浮动核电站空气中放射性核素扩散的研究

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-09-01 DOI: 10.1155/2021/1299821

Shuliang Zou, Na Liu, B. Huang

{"title":"Study on Airborne Radionuclide Dispersion in Floating Nuclear Power Plant under the Loss-of-Coolant Accident","authors":"Shuliang Zou, Na Liu, B. Huang","doi":"10.1155/2021/1299821","DOIUrl":"https://doi.org/10.1155/2021/1299821","url":null,"abstract":"Floating nuclear power plant is a kind of nuclear power plant on a barge moored specifically in an area of the sea. In order to study the factors influencing airborne radionuclide dispersion induced by the loss-of-coolant accident in floating nuclear power plant, the floating nuclear power plant platform was taken as the research object, and the dispersion of airborne radionuclide under combined conditions of platform positions, wind directions, and break directions (north, south, west, and east) was simulated by the CFD (computational fluid dynamics) method. The results show that northern and southern breaks have less dangerous island area than western and eastern ones but have more platform dangerous area than the western and eastern ones. The risk of the southern break is the greatest, and that of the western break is the least. Rotating the floating nuclear power plant platform in a certain angle can reduce the damage of loss-of-coolant accident. The effects of the dose received by the personnel under the condition of the severe accident were evaluated based on previous research, showing that the inhalation effective dose and the effective dose of plume immersion exposure were less than the radiation dose limit of 0.25 Sv within two hours in the accident. The results of the study can provide reference for the design of floating nuclear power plant platform and the formulation of emergency plan.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":"1 1","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"41374830","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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: 2

Numerical Investigation on Turbulent Flow and Heat Transfer of Helium-Xenon Gas Mixture in a Circular Tube 氦氙混合气体在圆管内湍流和传热的数值研究

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-08-30 DOI: 10.1155/2021/8356893

Biao Zhou, Han Zhang, Yu Ji, Jun Sun, Yuliang Sun

{"title":"Numerical Investigation on Turbulent Flow and Heat Transfer of Helium-Xenon Gas Mixture in a Circular Tube","authors":"Biao Zhou, Han Zhang, Yu Ji, Jun Sun, Yuliang Sun","doi":"10.1155/2021/8356893","DOIUrl":"https://doi.org/10.1155/2021/8356893","url":null,"abstract":"Gas-cooled space nuclear reactor system usually utilizes the helium-xenon gas mixture as the working fluid. Since the typical helium-xenon mixture has the Prandtl number of about 0.2, which is lower than that of water and air, the turbulent flow and heat transfer features need to be further investigated among the helium-xenon mixture and other fluids. In the current paper, numerical investigations by ANSYS Fluent are performed on helium-xenon mixture flow (HeXe40, M = 40.0 g/mol, Pr = 0.21), airflow (Pr = 0.71), and water flow (Pr = 6.99) in the circular tube. Direct numerical simulation results of liquid metal flow (Pr = 0.01) are also adopted for comparison. Results show that the dimensionless velocity profile and shear stress in the boundary layer of HeXe40 are close to those of other fluids. The empirical correlations from other fluids can also predict well the friction factor of helium-xenon mixtures. Due to the discrepancy in turbulent heat diffusivity ratio, the dimensionless radial temperature profile and turbulent heat conduction of HeXe40 significantly differ from those of other fluids. The molecular conduction region of HeXe40 develops up to y+ ≈ 30 and extends to the logarithmic region of the flow boundary layer. Moreover, the available experimental Nusselt numbers of helium-xenon mixtures are compared with several convective heat transfer correlations, in which Kays correlation is better.","PeriodicalId":21629,"journal":{"name":"Science and Technology of Nuclear Installations","volume":" ","pages":""},"PeriodicalIF":1.1,"publicationDate":"2021-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47775712","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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: 2

Studying Disturbance Wave Velocity and Wall Shear Stress of Vertical Upward Annular Flow in Narrow Rectangular Channel 窄矩形通道中垂直向上环形流动的扰动波速和壁面剪应力研究

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-08-27 DOI: 10.1155/2021/1927479

Antai Liu, Chang-qi Yan, Fuqiang Zhu, Haifeng Gu, Suijun Gong

引用次数: 0

A Correlation-Based Feature Selection Algorithm for Operating Data of Nuclear Power Plants 基于相关性的核电站运行数据特征选择算法

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-08-27 DOI: 10.1155/2021/9994340

Yuxuan He, Hongxing Yu, Ren Yu, Jian Song, Haibo Lian, Jiangyang He, Jiangtao Yuan

引用次数: 5

COMSOL Simulation for Design of Induction Heating System in VULCAN Facility VULCAN装置感应加热系统设计的COMSOL仿真

IF 1.1 4区工程技术

Science and Technology of Nuclear Installations Pub Date : 2021-08-19 DOI: 10.1155/2021/9922503

Jinkun Min, Guangyu Zhu, Yi Yuan, Jingquan Liu

引用次数: 2