核蒸汽系統 的英文怎麼說

中文拼音 [hézhēngqìxìtǒng]
核蒸汽系統 英文
nuclear steam system
  • 核 : 核構詞成分。
  • 蒸 : Ⅰ動詞1. （蒸發） evaporate2. （利用水蒸氣的熱力使食物熟或熱） steam Ⅱ名詞[中醫] （將藥物隔水蒸熟） steaming
  • 汽 : 名詞1. （由液體或某些固體變成的氣體） vapour2. （水蒸氣） steam
  • 系 : 系動詞（打結; 扣） tie; fasten; do up; button up
  • 統 : Ⅰ名詞1 （事物間連續的關系） interconnected system 2 （衣服等的筒狀部分） any tube shaped part of ...
  • 蒸汽 : vapour; reek; breath; steam蒸汽採暖 steam heating; 蒸汽錘 [機械工程] steam hammer; 蒸汽打樁鍾 ste...
  • 系統 : 1. （按一定關系組成的同類事物） system 2. （有條理的;有系統的） systematic
  1. Presents the steady - state programming with constant coolant average temperature and constant live steam pressure, an ideal programming for marine nuclear power plant ( npp ), under which, both the coolant average temperature and the live steam pressure of nuclear steam supply system ( nsss ) are constant when the steady state operating condition is varied, and points out that in comparison with the steadysate programming with constant coolant average temperature, the ideal programming can improve the overall operating performance of npp effectively, especially in low - load range, for instance, raise the heat efficiency of npp, decrease the operating noise of main coolant pump, and improve the conditions of design, operation and control of the secondary loop system

    雙恆定運行方案是船用壓水堆核動力裝置的理想運行方案.在這種運行方案下，當裝置穩態功率變化時，冷卻劑平均溫度與蒸汽壓都保持不變，與通常採用的冷卻劑平均溫度恆定運行方案相比，有效地改善了核動力裝置的總體運行性能，特別是在低負荷運行時，可以提高裝置經濟性、降低主泵運行噪聲、改善二迴路系統的設計、運行和控制條件，這對于提高船用核動力裝置的運行可靠性和安全性具有實際意義
  2. Nuclear steam supply systems

    核裝置蒸汽供給系統
  3. Main steam safety valve is the protective device against overpressure in the steam system of nuclear island

    主蒸汽安全閥是壓水堆核電站核島內主蒸汽系統的超壓保護裝置。
  4. Condensation of steam in the presence of noncondensable gases is an important heat transfer mode in the passive containment cooling system ( pccs ). in the past decade, great efforts have been made on the investigation of this regime

    含有不凝性氣體的蒸汽冷凝是先進反應堆非能動安全殼冷卻系統（ pccs ）中重要的換熱方式，是當前核反應堆熱工水力學研究的一個熱點。
  5. Test investigation on sgtr of single tube break and three - tube break is carried out on hpitf of ciae to deeply understand the thermal - hydraulic behavior during pressurized - water - reactor ( pwr ) sgtr and to provide an integral data base for code assessment and development. the facility, a full height and 1 / 1000 - scale ( by volume ) of qinshan nuclear power plant, incorporates two identical loops with active pumps and steam generators ( sg ). the initial steady state of the test basically simulated that of the reference plant operation

    為了深入了解核電廠蒸汽發生器傳熱管破裂事故發生后的熱工水力現象，以及對壓水堆（ pwr ）核電廠系統分析程序進行分析和評價，在高壓綜合試驗裝置上開展了蒸汽發生器傳熱管破裂試驗研究高壓綜合試驗裝置以秦山核電廠主迴路系統為模擬對象，該裝置為雙環迴路系統，容積比為1 ： 1000 ，高度比為1 ： 1 。
  6. A hybrid - model engineer simulator for nuclear steam supply systems

    ：核能蒸汽供給系統的混合模型工程師模擬器
  7. The break was assumed to be a single tube double - ended and three - tube double - ended shear break. the diameter of the break orifice is 0. 876mm and 1. 518mm respectively. the characteristic responses of main thermal - hydraulic parameters, such as primary pressure, secondary pressure, break flow rate, pressurizer level, secondary side level of steam generator and so on, were presented in this thesis

    試驗再現了核電廠蒸汽發生器單根傳熱管雙端斷裂事故和三根傳熱管雙端斷裂事故發生后的熱工水力現象，給出了事故過程中一次系統壓力、二次系統壓力、破口流量、蒸汽發生器二次側水位、穩壓器水位、元件表面溫度等主要參數的瞬態響應。
  8. Major : material science postgraduate : zhou xuchang advisor : zeng guangting huo yongzhong as a new type of shape memory alloy pipe coupling, nitinb pipe coupling has a good prospects on the connecting pipe. the study on the strength of nitinb connnetor system and design to optimize its structure have great meaning to its application on the connecting pipe of the steam maker in nuclear reactor on naval vessels

    Nitinb管接頭是一種新型形狀記憶合金管接頭，在管路連接上有著良好的應用前景。研究管接頭系統的強度並對其結構優化設計對它在艦艇核反應堆蒸汽發生器的管路連接上的應用有著重要的意義。
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