Steam generator tube rupture(SGTR) accident is an important scenario needed to be considered in the safety analysis of lead-based fast reactors. When the steam generator tube breaks close to the main pump, water vapor...Steam generator tube rupture(SGTR) accident is an important scenario needed to be considered in the safety analysis of lead-based fast reactors. When the steam generator tube breaks close to the main pump, water vapor will enter the reactor core, resulting in a two-phase flow of heavy liquid metal and water vapor in fuel assemblies. The thermal-hydraulic problems caused by the SGTR accident may seriously threaten reactor core's safety performance. In this paper, the open-source CFD calculation software OpenFOAM was used to encapsulate the improved Euler method into the self-developed solver LBEsteamEulerFoam. By changing different heating boundary conditions and inlet coolant types, the two-phase flow in the fuel assembly with different inlet gas content was simulated under various accident conditions. The calculation results show that the water vapor may accumulate in edge and corner channels. With the increase in inlet water vapor content, outlet coolant velocity increases gradually. When the inlet water vapor content is more than 15%, the outlet coolant temperature rises sharply with strong temperature fluctuation. When the inlet water vapor content is in the range of 5–20%, the upper part of the fuel assembly will gradually accumulate to form large bubbles. Compared with the VOF method, Euler method has higher computational efficiency. However, Euler method may cause an underestimation of the void fraction, so it still needs to be calibrated with future experimental data of the two-phase flow in fuel assembly.展开更多
This paper assessed the benefit of the in-situ pressure test to support steam generator tube integrity assessment and reviewed a conservatism of currently applied structural integrity assessment methodology against de...This paper assessed the benefit of the in-situ pressure test to support steam generator tube integrity assessment and reviewed a conservatism of currently applied structural integrity assessment methodology against defected tubes. According to the steam generator program requirement, condition monitoring assessment was performed to the all detected flaws. For condition monitoring assessments, the limiting structural integrity requirement should be demonstrated for all detected degraded tubes at a probability of at least 0.95 at 50% confidence. Some flaws were slightly exceeded the structural integrity threshold values of the condition monitoring performance limits using analytical method. As a direct evaluation of tube integrity with degraded tubes, in-situ pressure testing performed on some selected flaws and passed all proof and leakage test criteria with no leakage. From this pressure testing, the authors have verified that degraded tubes met a specified value containing a defined safety margins. Also, the authors have confirmed that existing structural assessment methodology has enough margins to retain integrity of steam generator tubes.展开更多
螺旋管蒸汽发生器是液态金属快堆中能量传递的核心设备,其运行的稳定性、安全性对核电站的运行有至关重要的影响。为此,本文构建了液态金属快堆螺旋管蒸汽发生器一次侧、二次侧耦合传热的三维数值模型,分别基于经济合作与发展组织核能署...螺旋管蒸汽发生器是液态金属快堆中能量传递的核心设备,其运行的稳定性、安全性对核电站的运行有至关重要的影响。为此,本文构建了液态金属快堆螺旋管蒸汽发生器一次侧、二次侧耦合传热的三维数值模型,分别基于经济合作与发展组织核能署(The Organisation for Economic Co-operation and Development,OECD/NEA)物性手册和美国国家标准与技术研究院(National Institute of Standards and Technology,NIST)数据库建立液态金属和水-水蒸气变物性计算关联式,采用Lee相变模型计算二次侧水-水蒸气蒸发过程中两相间的质量传递。基于实验数据,分别对本文模型一次侧传热以及二次侧传热的计算可靠性进行了验证。最后以铅铋快堆为例,研究了不同一次侧进口参数下蒸汽发生器一、二次侧之间的耦合传热特性,并与传统水冷堆进行了对比。结果表明:在同等条件下,相比于传统水冷堆,一次侧采用铅铋液态金属时,一、二次侧之间的壁面热流密度明显提升,热流密度峰值可达1439.97 kW·m^(-2),比水冷堆相应数值提升5~6倍,这导致二次侧管内气相蒸发过程明显加剧,体积含气率急剧上升;同时,一、二次侧之间的沿程热流密度分布更加不均匀,沿程热流密度分布相对偏差值比水冷堆相应数值增大3~4倍。随着一次侧进口铅铋温度从350℃增大到450℃,一、二次侧之间的壁面热流密度随之增大,对应的热流密度峰值从950.7 kW·m^(-2)增大到1439.97 kW·m^(-2),提升约1.5倍,同时一、二次侧之间的沿程热流密度分布更加不均匀,不均匀度增大20%。展开更多
基金supported partly by the Ministry of Science and Technology of the People's Republic of China (No. 2020YFB1902100)the Shanghai Municipal Commission of Economy and Informatization (No. GYQJ-2018-2-02)。
文摘Steam generator tube rupture(SGTR) accident is an important scenario needed to be considered in the safety analysis of lead-based fast reactors. When the steam generator tube breaks close to the main pump, water vapor will enter the reactor core, resulting in a two-phase flow of heavy liquid metal and water vapor in fuel assemblies. The thermal-hydraulic problems caused by the SGTR accident may seriously threaten reactor core's safety performance. In this paper, the open-source CFD calculation software OpenFOAM was used to encapsulate the improved Euler method into the self-developed solver LBEsteamEulerFoam. By changing different heating boundary conditions and inlet coolant types, the two-phase flow in the fuel assembly with different inlet gas content was simulated under various accident conditions. The calculation results show that the water vapor may accumulate in edge and corner channels. With the increase in inlet water vapor content, outlet coolant velocity increases gradually. When the inlet water vapor content is more than 15%, the outlet coolant temperature rises sharply with strong temperature fluctuation. When the inlet water vapor content is in the range of 5–20%, the upper part of the fuel assembly will gradually accumulate to form large bubbles. Compared with the VOF method, Euler method has higher computational efficiency. However, Euler method may cause an underestimation of the void fraction, so it still needs to be calibrated with future experimental data of the two-phase flow in fuel assembly.
文摘This paper assessed the benefit of the in-situ pressure test to support steam generator tube integrity assessment and reviewed a conservatism of currently applied structural integrity assessment methodology against defected tubes. According to the steam generator program requirement, condition monitoring assessment was performed to the all detected flaws. For condition monitoring assessments, the limiting structural integrity requirement should be demonstrated for all detected degraded tubes at a probability of at least 0.95 at 50% confidence. Some flaws were slightly exceeded the structural integrity threshold values of the condition monitoring performance limits using analytical method. As a direct evaluation of tube integrity with degraded tubes, in-situ pressure testing performed on some selected flaws and passed all proof and leakage test criteria with no leakage. From this pressure testing, the authors have verified that degraded tubes met a specified value containing a defined safety margins. Also, the authors have confirmed that existing structural assessment methodology has enough margins to retain integrity of steam generator tubes.
文摘螺旋管蒸汽发生器是液态金属快堆中能量传递的核心设备,其运行的稳定性、安全性对核电站的运行有至关重要的影响。为此,本文构建了液态金属快堆螺旋管蒸汽发生器一次侧、二次侧耦合传热的三维数值模型,分别基于经济合作与发展组织核能署(The Organisation for Economic Co-operation and Development,OECD/NEA)物性手册和美国国家标准与技术研究院(National Institute of Standards and Technology,NIST)数据库建立液态金属和水-水蒸气变物性计算关联式,采用Lee相变模型计算二次侧水-水蒸气蒸发过程中两相间的质量传递。基于实验数据,分别对本文模型一次侧传热以及二次侧传热的计算可靠性进行了验证。最后以铅铋快堆为例,研究了不同一次侧进口参数下蒸汽发生器一、二次侧之间的耦合传热特性,并与传统水冷堆进行了对比。结果表明:在同等条件下,相比于传统水冷堆,一次侧采用铅铋液态金属时,一、二次侧之间的壁面热流密度明显提升,热流密度峰值可达1439.97 kW·m^(-2),比水冷堆相应数值提升5~6倍,这导致二次侧管内气相蒸发过程明显加剧,体积含气率急剧上升;同时,一、二次侧之间的沿程热流密度分布更加不均匀,沿程热流密度分布相对偏差值比水冷堆相应数值增大3~4倍。随着一次侧进口铅铋温度从350℃增大到450℃,一、二次侧之间的壁面热流密度随之增大,对应的热流密度峰值从950.7 kW·m^(-2)增大到1439.97 kW·m^(-2),提升约1.5倍,同时一、二次侧之间的沿程热流密度分布更加不均匀,不均匀度增大20%。
