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堆堆芯
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  reactor core
     Aseismic Analysis of Pulsed Reactor Core Structure
     脉冲堆堆芯结构抗震分析
短句来源
     Single Assembly Preliminary Analysis for Horizontal Seismic Analysis on Fast Breeder Reactor Core
     快堆堆芯水平抗震分析的单组件初步分析
短句来源
     Real Time Simulation Research in 200 MW Low Temperature Nuclear Heating Reactor Core
     200MW低温供热堆堆芯实时仿真研究(英文)
短句来源
     By using lattice calculation code WIMS D/4,hexagonal nodal diffusion code SIXTUS and the WIMS N1/N2 library which include the data of the nuclide H in ZrH,Er 166 and Er 167,nuclear safety parameters are analyzed of uranium zirconium hydride pulsed reactor core,power peaking factor and fuel temperature coefficient.
     应用自己扩充的含有氢化锆中氢、铒166和铒167核素数据的WIMSN1/N2数据库以及国际通用的栅元计算程序WIMSD/4和六角形节块程序SIXTUS,分析了铀氢锆脉冲堆堆芯重要的核安全参数:功率峰因子和燃料负温度系数。
短句来源
     Thermal hydraulic design of reactor core for 10MW research reactor together with accepted criteria and design base is presented in this paper.
     简述了10MW研究堆堆芯热工水力设计的准则、设计基础和CTSA程序特点。
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  “堆堆芯”译为未确定词的双语例句
     PWR In-Core Fuel Management Package—PSUI-LEOPARD/ADMARC and PSUI-LEOPARD/NGMARC
     压水堆堆芯燃料管理软件包——PSUI-LEOPARD/ADMARC 和 PSUI-LEOPARD/NGMARC
短句来源
     Experimental Study on Measurement Method of Exit Temperature of Coolant of Fuel Assembly in Core of 200 MW Nuclear Heating Reactor
     200MW核供热堆堆芯燃料元件盒冷却剂出口温度测量方法实验研究
短句来源
     Dynamic Modeling of 10MW HTR Core and Its Application
     10MW高温气冷堆堆芯的动态建模及其应用
短句来源
     Some problems of aseismic test model design of NHR-200 core structure
     200MW供热堆堆芯结构抗震试验模型设计的若干问题
短句来源
     The method of frequency measurement by using spectrum analysis for the frequency of output signal of turbine flow transmitter is described and its application in flow measurement of the kernel flow of the 5MW low temperature nuclear heating pile is introduced.
     介绍频谱分析方法频率测量、测量涡轮流量变送器输出信号频率的方法及在5MW低温核供热堆堆芯流量测量中的应用.
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  相似匹配句对
     A STUDY ON THE CHOICE OF PARAMETERS OF TOKAMAK FUSION REACTOR CORE
     托卡马克聚变芯参数的选择
短句来源
     Aseismic Analysis of Pulsed Reactor Core Structure
     脉冲芯结构抗震分析
短句来源
     PLASMA BURNING AND CONTROL IN REACTOR CORE
     堆芯等离子体燃烧和控制
短句来源
     Core Design for AC-600
     AC—600堆芯设计
短句来源
     STUDY OF PLASMA BURNING AND CONTROL IN REACTOR
     堆芯等离子体燃烧和控制
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  reactor core
A new supercritical water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed.
      
By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained.
      
The possibility is discussed of using the obtained data in simulating emergency conditions in the reactor core.
      
The interaction of nuclear reactor core melt with oxide sacrificial material of localization device for a nuclear power plant wi
      
The neutron spectrum at the center of the reactor core was measured by the neutron activation method using the a priori spectrum as a superposition of partial fission and evaporation spectra.
      
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The authors had measured the transient of the thermal parameters(e.g.the inlet and outlet coolant temperature of fuel element case,the temperature offuel cladding,the flow of fuel element case and its thermal power)of the fuelelement at loss-of-power accident,the flow-reversal transient when emergencypump was starting and stopping during shutdown cooling,and the long-timedecay heat after shutdown by means of the in-core fuel element temperature-flowmeasurement facility of HFETR.The above-mentioned measurement...

The authors had measured the transient of the thermal parameters(e.g.the inlet and outlet coolant temperature of fuel element case,the temperature offuel cladding,the flow of fuel element case and its thermal power)of the fuelelement at loss-of-power accident,the flow-reversal transient when emergencypump was starting and stopping during shutdown cooling,and the long-timedecay heat after shutdown by means of the in-core fuel element temperature-flowmeasurement facility of HFETR.The above-mentioned measurement results havebeen described.

作者应用高通量工程试验堆堆芯燃料元件温度-流量测量装置测定了在全厂断电事故情况下的燃料元件热工参数(元件盒进出口水温,元件包壳温度,元件盒流量及其热功率)的瞬态过程,测定了在停堆冷却过程中启停事故泵时的流动反向过程,进行了停堆后的长时间剩余发热测量,给出了上述测量结果。

During the last decade,considerable progress has been in developing coarsemesh methods and modern nodal methods for solving the three-dimensional neutrondiffusion problems.However,to apply these methods successfully in the fuel management calcula-tions for the realistic light water reactor,we must solve the following problems:(1)Calculation of the nodal(assembly)homogenization parameters;(2)Calcul-ation of the detailed power distribution and the local quantities in an assembly;(3)Account explicitly for the local...

During the last decade,considerable progress has been in developing coarsemesh methods and modern nodal methods for solving the three-dimensional neutrondiffusion problems.However,to apply these methods successfully in the fuel management calcula-tions for the realistic light water reactor,we must solve the following problems:(1)Calculation of the nodal(assembly)homogenization parameters;(2)Calcul-ation of the detailed power distribution and the local quantities in an assembly;(3)Account explicitly for the local cross section variation caused by the deple-tion and the power dependent feedback.In this paper,we will review the recentresearch development in these areas.

近几年发展的粗网法和现代节块法应用于水堆堆芯燃料管理计算,必须解决以下几个问题:①节块(组件)均匀化参数的计算;②组件内精细功率分布和局部量的计算;③由于燃耗和功率反馈效应等所引起的局部截面变化的考虑。本文将评述这些问题的最近国内、外研究进展.

This article outlines a model for calculating the parameters oi a tokamak fusion reactor core. By numerical calculation and sensitivity analysis of uncertain factors in design, a reasonable set of parameters for the reactor core can be obtained under given conditions. When this model and calculation are applied to INTOR, the results are close to the published values.

本文概述了托卡马克聚变堆堆芯参数计算的一种基本模型。通过计算和不定因素对设计灵敏度的分析,得到了在一定条件下较为合理的一组堆芯参数。用该模型和程序对INTOR装置进行计算,所得结果与目前公布值接近。

 
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