Thermal-Hydraulic Analysis for Main Steam Line Break for Containment Pressure and Radioactivity Suppression System (CPRSS) of SMART Reactor

Kyung Jun Kang; Ji Han Chun; Jong Wook Kim; Han-Ok Kang; Seong Su Jeon; Keung Ku Kim; Moon Jung; Young-In Kim

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Thermal-Hydraulic Analysis for Main Steam Line Break for Containment Pressure and Radioactivity Suppression System (CPRSS) of SMART Reactor

机译：SMART反应堆安全壳压力和放射性抑制系统（CPRSS）主蒸汽管线断裂的热工水力分析

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来源
《Transactions of the American nuclear society》 |2019年第11期|1799-1799|共1页
作者
Kyung Jun Kang; Ji Han Chun; Jong Wook Kim; Han-Ok Kang; Seong Su Jeon; Keung Ku Kim; Moon Jung; Young-In Kim;
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SMART Development Institute Korea Atomic Energy Research Institute 989-111 Daedoek-daero Yuseong-gu Daejeon;

FNC Technology Co. 13 Heungdeok 1-ro Giheung-gu Yongin-si Gyeonggi-do Korea;

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入库时间 2022-08-18 05:24:42

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专利

1. Analysis for Pressure Oscillation Phenomena Induced by Steam Condensation in Containment with Pressure Suppression System, (I)Model and Analysis for Experimental Conditions of Marviken Reactor [J] . Motoaki OKAZAKI Journal of nuclear science and technology . 1979,第1期

机译：压力抑制系统中密闭蒸汽凝结引起的压力振荡现象的分析，（I）Marviken反应堆模型和实验条件分析
2. Simulation of a main steam line break accident using a coupled 'system thermal-hydraulics, three-dimensional reactor kinetics, and hot channel analysis' code [J] . Jae-Jun Jeong, Won Jae Lee, Bub Dong Chung Annals of nuclear energy . 2006,第9期

机译：使用耦合的“系统热工液压，三维反应堆动力学和热通道分析”代码模拟主蒸汽管线断裂事故
3. Qualification of Coupled 3-D Neutron-Kinetic/Thermal-Hydraulic Code Systems by the Calculation of Main-Steam-Line-Break Benchmarks in an NPP with WER-440 Reactor [J] . S. Kliem, S. Danilin, A. Haemaelaeinen, Nuclear science and engineering . 2007,第3期

机译：通过计算WER-440反应堆中核电厂的主蒸汽管道断裂基准，对耦合的3-D中子动力学/热工液压代码系统进行鉴定
4. A THERMAL-HYDRAULICS ANALYSIS OF PIPE BREAKS IN THE PEBBLE BED MODULAR REACTOR MAIN POWER SYSTEM [C] . Naison Mazana International conference on nuclear engineering;ICONE-16 . 2008

机译：卵床模块化反应堆主电力系统管道破裂的热力学分析。
5. Experimental study on suppression chamber thermal-hydraulic behavior for long-term reactor core isolation cooling system operation [D] . Solom, Matthew Alan. 2016

机译：长期反应器芯隔离冷却系统运行的抑制室热液压行为的实验研究
6. A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings [O] . Jianguo Zhou, Yaming Xu, Tao Zhang 2016

机译：反应堆安全壳表面裂缝无线监测系统
7. ANALYSIS OF THE OECD MAIN STEAM LINE BREAK BENCHMARK PROBLEM USING THE REFINED CORE THERMAL-HYDRAULIC NODALIZATION FEATURE OF THE MARS0MASTER CODE [O] . Han Gyu Joo, Jae-jun Jeong, Byung-oh Cho, 2001

机译：利用maRs0masTER代码的精炼核心热液化节点特征分析经合组织主要蒸汽管线断裂基准问题
8. Thermal-Hydraulic Analysis of Main-Steam-Line-Break Accidents as Potential Initiators for Reactor Vessel [R] . Lamkin, D. E. 1983

机译：作为反应堆容器潜在引发剂的主蒸汽管道断裂事故的热工水力分析

Thermal-Hydraulic Analysis for Main Steam Line Break for Containment Pressure and Radioactivity Suppression System (CPRSS) of SMART Reactor

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