coolant systems造句

• Dimensions and tolerances for coolant system hoses
  制冷系统软管尺寸和公差
• Nuclear reactor and primary coolant system
  核反应堆及主冷却剂系统
• Primary coolant system
  主冷却剂系统
• coolant system hoses
  制冷系统软管
• Main coolant system
  主冷却剂系统
• The real-time simulation of reactor coolant system was studies in the dissertation
  本论文对反应堆冷却剂系统进行了实时仿真计算研究。
• After testifying of equipment models, a one-loop coolant system is established and the model is tested
  对于稳压器，根据其内部特征，采用了三区模型。
• Nuclear reactors; instrumentation and control systems important for safety; detection of leakage in coolant systems
  核反应堆.安全控制装置.冷却设备中核泄漏检测
• Applications coolant systems, lubricant systems, detecting if material lacking, warning if pipelines jammingc
  用途：如冷却系统润滑系统缺料检测管路阻塞的紧告。
• Integral spindle coolant system decreases the high temperature and ensures running accuracy after long time working
  采用内藏式主轴油冷却系统，有效降低头部温升，确保长时间加工的运转精度。
• As a important part of the nuclear power plant simulation, reactor coolant system simulation research is of great significance
  作为核动力装置培训模拟器的重要组成部分，冷却剂系统的仿真计算研究具有很重要的意义。
• The primary coolant system of a 900 mw class reactor consists of the reactor pressure vessel and the primary circuit with 3 identical loop top figure
  一个900兆瓦的反应堆主冷却剂系统上图是由压力壳及三个相同的环路相连组成即一回路。
• The primary coolant system of a 900 mw class reactor consists of the reactor pressure vessel and the primary circuit with 3 identical loop ( top figure )
  一个900兆瓦的反应堆主冷却剂系统(上图)是由压力壳及三个相同的环路相连组成(即一回路)。
• In this paper, the method for transient characteristic calculation of pwr ( pressurized water reactor ) coolant system is studied and a model is developed
  本文根据压水堆冷却剂系统的特点，对反应堆、蒸汽发生器、稳压器分别建立各自独立的模型。
• In order to adapt to real time simulation, the nodes of system and the predigestion was made according to the characteristic of reactor coolant system which will be simulated
  本论文针对要仿真的冷却剂系统的特点，对系统进行了节点划分，并做了适当简化，以使其适应实时仿真的需要。
• The reactor coolant system is an important system of nuclear power plant, so simulation of reactor coolant system is an important part of nuclear power plant simulation
  主冷却剂系统是核动力装置的一个重要系统，对主冷却剂系统的仿真研究是核动力装置仿真研究的一个重要部分，具有非常重要的意义。
• The reactor coolant system is an important system of nuclear power plant, so simulation of reactor coolant system is an important part of nuclear power plant simulation
  主冷却剂系统是核动力装置的一个重要系统，对主冷却剂系统的仿真研究是核动力装置仿真研究的一个重要部分，具有非常重要的意义。
• A mathematic model of reactor, steam generator, pressurizer is developed based on characteristic of pwr coolant system . for reactor, the point reactor kinetics equation with six-group delayed neutrons is used and influence of coolant's and fuel's temperature is considered, conduction model of one dimension and single-channel model is used in thermal and hydraulic calculation, third-order hermite polynomial is used to solve the point reactor kinetics equation
  对于反应堆，在物理计算中采用了带有六组缓发中子的点堆方程，同时考虑了多普勒效应及冷却剂温度效应对反应性的影响；在热工水力计算中采用了一维的热传导模型与单通道模型；在求解点堆方程中采用了hermite插值的方法，克服了点堆方程的刚性问题。
• At the main content of the thesis, the author collect the main fault's character information of primary loop systems, and those systems include main coolant system and pressure safety system and redundant heat discharging system and safety injection system and so on, the author also compare those faults whose character information is similar
  在论文的主要工作内容中，本文收集整理了船用核动力装置一回路各系统的主要故障的特征信息，其中包括一回路主系统主冷却剂系统以及主要辅助系统如压力安全系统、余热排出系统、安全注射系统等，并对特征信息类似的故障进行了对比。
• In the thesis, the relap5 code is made use of simulating and calculating the passive core makeup tank ( cmt ) of the next generation reactor ac-600 of our country . and the calculative results based on the comparative character of the system are compared with the experimental results in the corresponding experimental devices of ac-600 makeup water tank in nuclear power institute of china ( npic ) . finally the thesis conclusion summarized is that the makeup water passively injected to the reactor coolant system by the core makeup tank depending on the gravity is viable in the middle and small locas
  本论文利用relap5程序对我国下一代先进堆ac-600的一个具有非能动性能的堆芯补水箱分系统进行了仿真计算，根据系统之间的可比性，把计算的结果和与之相对应的中国核动力研究设计院堆芯补水箱补水实验装置的实验结果进行了定性的比较分析，得到了论文的结论：在中、小破口失水事故的工况下，堆芯补水箱依靠重力向反应堆冷却剂系统实现非能动的补水是可行的。