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Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

机译：核反应堆的热水水力分析因丧失冷却剂事故，无需紧急核心冷却系统

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Pronob Deb Nath; Kazi Mostafijur Rahman; Md. Abdullah Al Bari;
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1. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1 [J] . Seon Oh Yu, Yong Jin Cho, Sung Joong Kim Nuclear engineering and technology . 2017,第5期

机译：Wolsong 1号机组大破口失水事故在再循环阶段应急堆芯冷却系统流量减少对通道温度的影响
2. Thermal analysis of prismatic gas-cooled reactor core under coolant channel blockage accidents [J] . Sung Nam Lee, Nam-il Tak, Min Hwan Kim, Annals of nuclear energy . 2014,第sepa期

机译：棱柱形气冷堆芯在冷却液通道堵塞事故下的热分析
3. Substantiating the Reliability of Cooling the Core of VVER Reactors during Accidents Involving Loss of Tightness of the Reactor Coolant Circuit [J] . B. F. Balunov, A. S. Babykin, A. A. Falkov, Thermal engineering . 2008,第1期

机译：证实在涉及反应堆冷却剂回路密封性损失的事故中冷却VVER反应堆堆芯的可靠性
4. AN EXPERIMENTAL INVESTIGATION ON THERMAL-HYDRAULIC INTERACTION BETWEEN CORE MAKEUP TANK AND SAFETY INJECTION TANK OF AN INTEGRAL SMALL REACTOR DURING A SMALL BREAK LOSS-OF-COOLANT ACCIDENT [C] . Hwang Bae, Woo Shik Kim, Sung Uk Ryu, ASME/JSME/KSME Joint Fluids Engineering Conference . 2016

机译：小型休闲冷却剂事故中整体小型反应器中芯构罐与安全注射槽热液压相互作用的实验研究
5. Probabilistic risk assessment for a loss of coolant accident in McMaster Nuclear Reactor and application of reliability physics model for modeling human reliability. [D] . Ha, Taesung. 2007

机译：McMaster核反应堆中冷却剂损失事故的概率风险评估，以及用于建立人类可靠性模型的可靠性物理模型的应用。
6. Development of an automatedchemistry control system forsecondary coolant circuits inCANDU nuclear power reactors [O] . J. R. Dean, R. B. Stewart 1979

机译：开发自动化化学控制系统二次冷却液回路坎杜核电反应堆
7. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR) [O] . Ronaldo Henriques Szilard 2016

机译：冷却液事故（LOCA）/紧急核心冷却液系统（ECCS对被告知的利润率的评估为代表加压水反应堆（PWR）
8. ECCSA-I A DIGITAL COMPUTER PROGRAM FOR THERMAL AND HYDRAULIC ANALYSIS OF CORE CHANNELS IN THE EVENT OF A NUCLEAR-REACTOR LOSS-OF-COOLANT ACCIDENT [R] . Jay S. Iyer 1968

机译：ECCsa-I核反应堆损失事故中核心通道的热量和液压分析的数字计算机程序

Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

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