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组件堆芯
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  assembly core
     THE APPLICATION OF THE FINITE ELEMENT NUMERICAL METHOD TO CRITICAL PROBLEM OF TWO DIMENSIONAL HEXAGONAL ASSEMBLY CORE
     有限元方法在二维六角形组件堆芯临界问题计算中的应用
短句来源
     The critical calculation of Few group neutron diffusion problem for Two dimensional Hexagonal assembly core Using finite element numerical method is studied.
     应用有限元方法对二维六角形组件堆芯的少群中子扩散临界计算进行了研究,并编制了二维少群中子扩散方程有限元计算程序2DFME。 以快堆300MW(e)-LMFBR临界问题[2]为例,作了数值计算。
短句来源
  “组件堆芯”译为未确定词的双语例句
     PRELIMINARY STUDY OF PHYSICAL PROPERTIES FOR HEXAGON LATTICE PWR CORE USING THORIUM
     压水堆加钍(ThO_2)六角形组件堆芯物理特性的初步探讨
短句来源
     Application Research for VVER Fuel Management Code System
     六角形组件堆芯燃料管理程序包的应用开发
短句来源
     So it is necessary to develop the VVER fuel management code packages.
     随着田湾核电站两台机组即将投产运行,六角形组件堆芯燃料管理程序包的研制开发十分重要。
短句来源
     Compared with full uranium fuel assemblies,more than 200 kg 235 U could be saved for each cycle by using separaed U Th assemblies. It shows the prosperous future of Th U fuel cycle.
     并可同全铀组件堆芯比较中看出 ,分立型铀、钍组件混装堆芯每一循环 (第 1 0循环后 )可少装 2 0 0多kg2 3 5U ,这样就为钍 铀燃料循环展示了光明的前景。
短句来源
     In this paper, the disturbance to the neutron field caused by the appearance of different backing materials is studied in detail with Monte-Carlo methods, which helps us in choosing the backing material reasonably and the data analysis in neutron flux measurements.
     本工作采用蒙特卡罗模拟计算办法,分析了多种衬底材料在不同厚度及不同衬底方式下对中子场的扰动情况,为板型燃料组件堆芯中子注量率测量中衬底材料的选择及测量数据的处理提供了依据。
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  相似匹配句对
     Plate type fuel assemblies are adopted in the core.
     堆芯采用平板式燃料组件
短句来源
     Seismic response analysis of fuel assembly in nuclear power station
     核电厂堆芯燃料组件地震反应分析
短句来源
     modular;
     组件模块化;
短句来源
     NET, and a graphical .
     NET组件
短句来源
     Core Design for AC-600
     AC—600堆芯设计
短句来源
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  assembly core
The characters of self-assembly core/shell nanoparticles of amphiphilic hyperbranched polyethers as drug carriers
      
The characters of self-assembly core/shell nanoparticles of amphiphilic hyperbranched polyethers (HP-g-PEO) as drug carriers were investigated.
      
The neutron spectra at the center of the critical assembly core and in the analog BR-1 reactor are compared.
      


The critical calculation of Few group neutron diffusion problem for Two dimensional Hexagonal assembly core Using finite element numerical method is studied.

应用有限元方法对二维六角形组件堆芯的少群中子扩散临界计算进行了研究,并编制了二维少群中子扩散方程有限元计算程序2DFME。以快堆300MW(e)-LMFBR临界问题[2]为例,作了数值计算。且与国内外有限元程序[1~3]和有限差分程序CITATION的计算结果进行了分析比较。

A practical application of 232 Th in Qinshan 300 MW(e) PWR was searched by using separated U Th fuel assemblies and the Twin Refuelling system.From the calculation of ten cycles,it shows that if irradiation time of thorium assemblies in core of PWR is long enough (10 cycles),the amount of 233 U in core at the end of irradiation is increased to 212.6 kg and it could take part in the chain reaction of the core.In this way,the practical application of 232 Th could be reached.Compared...

A practical application of 232 Th in Qinshan 300 MW(e) PWR was searched by using separated U Th fuel assemblies and the Twin Refuelling system.From the calculation of ten cycles,it shows that if irradiation time of thorium assemblies in core of PWR is long enough (10 cycles),the amount of 233 U in core at the end of irradiation is increased to 212.6 kg and it could take part in the chain reaction of the core.In this way,the practical application of 232 Th could be reached.Compared with full uranium fuel assemblies,more than 200 kg 235 U could be saved for each cycle by using separaed U Th assemblies.It shows the prosperous future of Th U fuel cycle.Of course,there are still some engineering problems left to be solved for real application.

通过对分立型铀、钍燃料组件 ,使用在秦山 30 0MW电功率压水堆核电厂中堆芯物理特性的探讨 ,寻找2 3 2 Th在PWR中可能利用的途径。为此 ,特采用铀、钍燃料组件分立的双进料系统的装卸料方法 ,其堆芯寿期分别为铀组件 3个循环 ;钍组件 1 0个循环。并以秦山核电厂为参考电厂 ,进行了 1 0个循环的燃耗计算 ,每一循环装料时均有 4个钍组件进堆。计算结果表明 :到第 1 0循环寿期末 ,堆芯中 40个钍组件所含的2 3 3 U总量已达到 2 1 2 6kg ,可直接参与堆芯的链式反应 ,从而达到利用2 3 2 Th的目的。并可同全铀组件堆芯比较中看出 ,分立型铀、钍组件混装堆芯每一循环 (第 1 0循环后 )可少装 2 0 0多kg2 3 5U ,这样就为钍 铀燃料循环展示了光明的前景。当然如果要达到实际应用 ,仍有许多工程技术问题亟待解决

In the course of the core flux density measurement, detector foil and its backing materials are introduced into the gap between slab-type fuel element. Thus the moderale medium is changed and as well as the moderation and absoration of neutrons. It brings error in measurement results, which should been considered. In this paper, the disturbance to the neutron field caused by the appearance of different backing materials is studied in detail with Monte-Carlo methods, which helps us in choosing the backing material...

In the course of the core flux density measurement, detector foil and its backing materials are introduced into the gap between slab-type fuel element. Thus the moderale medium is changed and as well as the moderation and absoration of neutrons. It brings error in measurement results, which should been considered. In this paper, the disturbance to the neutron field caused by the appearance of different backing materials is studied in detail with Monte-Carlo methods, which helps us in choosing the backing material reasonably and the data analysis in neutron flux measurements.

板型燃料组件窄流道内的中子处于欠慢化状态,在堆芯中子注量率测量中,探测片衬底材料的引入将较大地改变窄流道内慢化介质的成分,从而影响中子的慢化和吸收,给测量结果带来较大误差。本工作采用蒙特卡罗模拟计算办法,分析了多种衬底材料在不同厚度及不同衬底方式下对中子场的扰动情况,为板型燃料组件堆芯中子注量率测量中衬底材料的选择及测量数据的处理提供了依据。

 
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