nuclear reactor core 中文意思是什麼

nuclear reactor core 解釋
核反應堆堆芯
  • nuclear : adj 1 核的,成核的;有核的。2 【物理學】原子核的;原子能的;原子彈的;核動力的。3 〈比喻〉核心的...
  • reactor : n. 1. 反應者,被試驗者;【醫學】有(陽性)反應的人[動物]。2. 【電學】礙圈,扼流圈;電抗器;【化學】反應器;【物理學】反應堆。
  • core : CORE =Congress of Racial Equality 〈美國〉爭取種族平等大會。n 1 果心。2 (事物、問題等的)中心,...
  1. Nuclear reactor instrumentation - pressurized water reactor of vver design - monitoring adequate cooling within the core during shutdown

    核反應堆儀器儀表. vver設計的壓水反應堆.在定堆期間監測堆芯的充分冷卻
  2. The reactor can be shutdown during emergency by cutting off the power supply to the control rod driving mechanism which then causes the control rods to drop down to the reactor core by gravity quickly and thereby stopping the nuclear fission immediately

    在需要緊急停堆時, ?須切斷控制棒驅動機械的電源,控制棒便會因地心吸力而快速下墜至反應堆堆芯,立即停止核裂變。
  3. The paper reports the model of the reactor core of the pwr nuclear power plant and established the suitable core physics - mathematics model for microcomputer simulation

    摘要闡述了pwr核電站堆芯的模型化問題,提出了適用於微機模擬的核電站堆芯的物理數學模型。
  4. In order to meet the needs for measuring the coolant water flow rate in - core of a nuclear heating reactor, two types of turbine flow meter with low rotation speed have been developed in our institute

    為適應核反應堆堆芯冷卻劑流量測量的需要,開發研製了新型低速渦輪流量變送器,按流量信號輸出不同,分別為磁感應模擬信號輸出和數字開關量輸出低速渦輪流量變送器。
  5. In - core temperature or primary envelope temperature measurements in nuclear power reactor - characteristics and test methods

    核動力堆堆芯或堆主包殼內溫度測量特性和測試方法
  6. 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

    本論文利用relap5程序對我國下一代先進堆ac - 600的一個具有非能動性能的堆芯補水箱分系統進行了模擬計算,根據系統之間的可比性,把計算的結果和與之相對應的中國核動力研究設計院堆芯補水箱補水實驗裝置的實驗結果進行了定性的比較分析,得到了論文的結論:在中、小破口失水事故的工況下,堆芯補水箱依靠重力向反應堆冷卻劑系統實現非能動的補水是可行的。
分享友人