共查询到19条相似文献,搜索用时 209 毫秒
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数值反应堆是基于先进耦合建模技术、大规模并行计算技术、先进的验证与确认(V&V)等技术,建立在超级计算机上,可实现实际反应堆各种物理过程高精细模拟预测的复杂软件系统。它是实际反应堆“外在”和“内在”的镜像,是先进的核反应堆设计优化、高效运行、事故预测和应急以及新材料研发等的试验验证平台。本文在简略综述国内外典型数值反应堆研究成果的基础上,描述了本课题组近期开发的数值反应堆核心软件组成体系,分析了数值反应堆对计算资源和存储资源的需求,并介绍了目前正在开展的中国数值反应堆原型系统(CVR1.0)的研究进展。进展主要包括:两相子通道热工水力模拟软件、单相CFD热工水力模拟软件、多尺度材料辐照损伤模拟软件、直接3D中子输运特征线法模拟软件,以及这些软件与欧美CASL、NEAMS、RPV等相关软件的对比和在神威、曙光等超级计算机上的测试结果。 相似文献
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《原子能科学技术》2019,(10)
数值反应堆是基于先进耦合建模技术、大规模并行计算技术、先进的验证与确认(VV)等技术,建立在超级计算机上,可实现实际反应堆各种物理过程高精细模拟预测的复杂软件系统。它是实际反应堆"外在"和"内在"的镜像,是先进的核反应堆设计优化、高效运行、事故预测和应急以及新材料研发等的试验验证平台。本文在简略综述国内外典型数值反应堆研究成果的基础上,描述了本课题组近期开发的数值反应堆核心软件组成体系,分析了数值反应堆对计算资源和存储资源的需求,并介绍了目前正在开展的中国数值反应堆原型系统(CVR1.0)的研究进展。进展主要包括:两相子通道热工水力模拟软件、单相CFD热工水力模拟软件、多尺度材料辐照损伤模拟软件、直接3D中子输运特征线法模拟软件,以及这些软件与欧美CASL、NEAMS、RPV等相关软件的对比和在神威、曙光等超级计算机上的测试结果。 相似文献
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核反应堆人误数据的收集 总被引:2,自引:0,他引:2
本文结合我国五座核反应堆人误数据收集的实践,讨论了核反应堆运行中人误数据收集的内容、要求、方法和质保措施等技术问题,并简述了这些反应堆人误数据的收集结果。 相似文献
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核反应堆反应性的测量技术的研究对于表征反应堆的安全运行和发挥其经济效益有着重要的意义。计算笔记将抽象、复杂的研究分析过程文字化、可视化。论文以核反应堆反应性的测量技术研究过程为例,详细阐述了其计算笔记的运用,并在此基础上建立标准化分析模型、进行设计改或设计优化,不仅有利于对研究过程的再认识,还有利于增强反应堆的反应性测量技术的核心技术竞争力。 相似文献
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日本小型核动力反应堆及其技术特点 总被引:2,自引:0,他引:2
日本原子能研究所研制了包括一体化船用堆(MRX)在内的几种小型核反应堆.MRX采用容器内置式控制棒驱动机构、水淹式安全壳、非能动余热排出系统;MR-100G和MR-1G是专门为区域供热和冷却系统提供能源,一回路系统自加压的全自然循环一体化压水堆.其排放物活性较低,小型化、模块式结构.可直接建于城市,甚至办公大楼的地下.,水下探测器用小型潜水反应堆(SCR)的设计思路与MRX基本相同.但一回路为全自然循环,日本小型核反应堆发展的技术思路清晰,注重用途的拓展,具有战略发展远见.在将我国大型核动力反应堆研制经验及其相应技术的推广方面,日本小型反应堆的发展思路值得借鉴。 相似文献
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本文系统介绍了“大型先进压水堆及高温气冷堆核电站”国家科技重大专项课题“CAP1400数值反应堆关键技术研究”的主要研究成果。课题首先分别开发了基于确定论方法和蒙特卡罗方法的高保真堆芯物理计算程序,然后开发了pin by pin先进子通道分析程序和基于精细网格的燃料棒性能分析程序,以此为基础建立了物理 热工 燃料性能多物理耦合的CAP1400数值反应堆系统。利用国际基准题VERA、AP1000启动物理实验参数对数值反应堆系统进行了验证和确认,并进一步实现了CAP1400大型先进压水堆的启动物理参数、循环模拟分析和部分功率能力分析的示范应用。数值结果表明,所开发的数值反应堆关键分析软件具有很高的计算精度,可直接服务于CAP1400的设计验证、物理启动和运行支持。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(10):780-785
A method is introduced to evaluate the degree of nuclear technology transfer; that is, the output powers of Japanese nuclear reactors constructed in these 20 years are chronologically plotted in a semi-log figure. All reactors plotted are classified into imported and domestic ones according to a value of domestication factor. A space between two historical trajectories of reactor construction may be interpreted as one of the measures indicating the degree of nuclear technology transfer. In connection with this method, historical change of educational and training courses in Nuclear Engineering School of Japan Atomic Energy Research Institute is reviewed in this report. 相似文献
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This paper reflects the thoughts and work concerning considerations and design of 200-MW nuclear heating reactor (NHR-200) developed in Institute of Nuclear Energy Technology (INET), Tsinghua University, China. Due to the fact that the size of heating reactors is limited to the local demands which are generally smaller that the economic reasonable size as compared to those reactors for electricity production, the design of systems for NHR-200 should be specified in accordance with its design characteristics, and simplified as much as possible for economic aim. The nuclear heating reactor has a low power density in the core and that the annual generation period is only about 180 days. Therefore, the total required number of fuel bundles is rather small. Furthermore, in-vessel spent fuel storage is feasible. All these features raise the potential to simplify the fuel storage system for NHR-200. The fuel storage and inspection facilities are described. 相似文献
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After the nuclear accidents of Three Mile Island and Chernobyl the world nuclear community made great efforts to increase research on nuclear reactors and to develop advanced nuclear power plants with much improved safety features. Following the successful construction and a most gratifying operation of the 10 MWth high-temperature gas-cooled test reactor (HTR-10), the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University has developed and designed an HTR demonstration plant, called the HTR-PM (high-temperature-reactor pebble-bed module). The design, having jointly been carried out with industry partners from China and in collaboration of experts worldwide, closely follows the design principles of the HTR-10.Due to intensive engineering and R&D efforts since 2001, the basic design of the HTR-PM has been finished while all main technical features have been fixed. A Preliminary Safety Analysis Report (PSAR) has been compiled.The HTR-PM plant will consist of two nuclear steam supply system (NSSS), so called modules, each one comprising of a single zone 250 MWth pebble-bed modular reactor and a steam generator. The two NSSS modules feed one steam turbine and generate an electric power of 210 MW.A pilot fuel production line will be built to fabricate 300,000 pebble fuel elements per year. This line is closely based on the technology of the HTR-10 fuel production line.The main goals of the project are two-fold. Firstly, the economic competitiveness of commercial HTR-PM plants shall be demonstrated. Secondly, it shall be shown that HTR-PM plants do not need accident management procedures and will not require any need for offsite emergency measures.According to the current schedule of the project the completion date of the demonstration plant will be around 2013. The reactor site has been evaluated and approved; the procurement of long-lead components has already been started.After the successful operation of the demonstration plant, commercial HTR-PM plants are expected to be built at the same site. These plants will comprise many NSSS modules and, correspondingly, a larger turbine. 相似文献
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E.L. Gluekler C.C. Bigelow V. DeVita J.M. Kelly R.W. Seidensticker F.F. Tajirian 《Nuclear Engineering and Design》1991,127(3)
GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US Department of Energy (DOE). The objective of this development is to provide, by the turn of the century, a reactor concept with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactors and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long-duration earthquakes. Seismic isolation is accomplished with high-damping natural-rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity.This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program, which is supported by the US Department of Energy (DOE), Argonne National Laboratory (ANL), Energy Technology Engineering Center (ETEC), the University of California at Berkeley (UC-Berkeley), General Electric (GE), and Bechtel National, Inc. (BNI), is described in this paper and selected results are presented. The initial testing indicated excellent performance of high-damping natural-rubber bearings. The development of seismic isolation guidelines is in progress as a joint activity between ENEA of Italy and the GE Team. 相似文献
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Matthieu Lemaire Hyunsuk Lee Nam-il Tak Hyun Chul Lee 《Journal of Nuclear Science and Technology》2017,54(6):668-680
This paper presents the application results of MCS/GAMMA+ to multi-physics analysis of OECD/NEA modular high temperature gas-cooled reactor (MHTGR) benchmark Phase I Exercise 3. It is a part of international R&D efforts lead by the Next Generation Nuclear Plant (NGNP) US project to improve the neutron-physics and thermal-fluid simulation of (high temperature gas-cooled reactors) HTGRs, one of the next generations of safer nuclear reactors. Accurate and validated analysis tools are indeed a crucial requirement for safety analysis and licensing of nuclear reactors. To guide this effort, a numerical benchmark on the MHTGR was created by the NGNP project and formally approved in 2012 for international participation by the OECD/NEA. The benchmark defines a common set of exercises and the comparison of solutions obtained with different analysis tools is expected to improve the understanding of simulation methods for HTGRs. The coupled neutronics/thermal-fluid solution presented in this paper was obtained with the neutron transport Monte Carlo code MCS developed by Ulsan National Institute of Science and Technology and the thermal-fluid code GAMMA+ developed by Korean Atomic Energy Research Institute. The purpose of this paper is to present the GAMMA+/MCS coupled system, the calculation methodology, and the obtained solutions. 相似文献
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核动力系统模拟技术的研究 总被引:2,自引:1,他引:1
简要回顾了清华大学核研院在系统模拟技术方面所开展的主要工作,重点介绍了基于RETRAN-02程序研究开发的200MW核供热堆紧凑型模拟器和基于网络计算技术的开发的10MW高温气冷堆网络并行模拟原型系统。 相似文献
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基于高保真模型和方法的数值反应堆具有高精度和高分辨率的特点,但核数据等参数固有的不确定度将严重影响其分析结果的不确定度。本文在综述国内外数值反应堆及其不确定度量化研究进展的基础上,重点介绍了西安交通大学核工程计算物理(NECP)实验室近年来在基于一步法的高保真数值反应堆程序NECP-X的研发与验证、核数据协方差数据库制作、基于确定论和抽样方法的不确定度传递方法研究及程序开发、核数据(包括截面、瞬发裂变谱、散射角分布等)的不确定度传递以及时空瞬态计算中的不确定度量化等方面的研究进展,提出了COST先进抽样方法,并首次基于高保真数值反应堆程序量化了各类核参数的协方差在堆芯稳态和瞬态分析中的不确定度传递,对于数值反应堆的工程应用具有重要意义。 相似文献
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Patrícia A.L. Reis Antonella L. Costa Cláubia Pereira Maria A.F. Veloso Amir Z. Mesquita Humberto V. Soares Graiciany de P. Barros 《Annals of Nuclear Energy》2010
RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations. 相似文献