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魏承君,陈子佳,庞思敏,等. AP1000核电站蒸汽发生器主给水管道双端断裂事故下第一跨空间三维流动特性数值模拟[J]. 科学技术与工程, 2021, 21(4): 1388-1393.
WEI Chengjun,et al.Numerical simulation on three-dimensional flow characteristics of steam generator main feedwater pipeline double-end fracture accident in the first span of AP1000[J].Science Technology and Engineering,2021,21(4):1388-1393.
AP1000核电站蒸汽发生器主给水管道双端断裂事故下第一跨空间三维流动特性数值模拟
Numerical simulation on three-dimensional flow characteristics of steam generator main feedwater pipeline double-end fracture accident in the first span of AP1000
投稿时间:2020-04-05  修订日期:2020-08-13
DOI:
中文关键词:  第一跨空间  蒸汽发生器给水管道断裂  防水淹计算  VOF  AP1000
英文关键词:the first span  the broken steam generator water supply pipe  waterproof calculation  VOF  AP1000
基金项目:国家科技重大专项(2018ZX06001001)第一作者:魏承君(1985—),男,汉族,山东潍坊人,学士,副高级工程师。研究方向:核电厂常规岛设计。E-mail:weichengjun@ snpdri.com。*通讯作者:张钰浩(1990—),男,汉族,山东滕州人,博士,讲师。研究方向:核电厂热工水力学。E-mail:zhangyuhao@ ncepu.edu.cn。 陈子佳2, 3 庞思敏1 赵海琦2, 3 张钰浩2, 3*
              
作者单位
魏承君 国核电力规划设计研究院有限公司
陈子佳 非能动核能安全技术北京市重点实验室
华北电力大学核科学与工程学院
庞思敏 国核电力规划设计研究院有限公司
赵海琦 非能动核能安全技术北京市重点实验室
华北电力大学核科学与工程学院
张钰浩 非能动核能安全技术北京市重点实验室
华北电力大学核科学与工程学院
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中文摘要:
      AP1000核电厂第一跨空间内布置了设备冷却水系统(CCS)驱动泵,能够保证核电厂事故工况下设备冷却水系统、余热排出系统等关键安全系统的正常运行,从而保证核电厂安全。然而在蒸汽发生器主给水管道双端断裂事故下,大量的水会泄放到第一跨空间内,对第一跨空间内的关键设备造成严重威胁。因此,本文对AP1000核电站蒸汽发生器主给水管道双端断裂事故下第一跨空间内泄放流体三维流动特性进行数值模拟研究。采用ANSYS系列软件,建立第一跨空间三维模型,基于两相流VOF模型,计算冷却剂喷放事故下,第一跨空间内流动特性及水位变化规律。计算结果表明,破口水从入口进入第一跨空间后在5. 334 m层漫流,绝大部分泄放水通过该层设置的预留开孔流出,部分水在该层堆积。但是,由于设置挡水沿,泄洪水并未漫流到0 m层与-3.8 m层,随着冷却剂喷放引发给水泵跳泵,第一跨空间内水位将逐渐下降,不会造成重要设备防水台的漫流淹没。计算结果对核电厂主要泄洪途径、关键设备布置设计与优化提供了数值参考。
英文摘要:
      In the first span of the AP1000 nuclear power plant, the cooling water system (CCS) drive pumps are arranged, which can ensure the normal operation of key safety systems such as equipment cooling water systems and waste heat removal systems under nuclear power plant accident conditions, thereby ensuring nuclear power plant safety. However, in the event of a double-ended fracture of the main feedwater pipe of the steam generator, a large amount of water will be discharged into the first span space, posing a serious threat to the key equipment in the first span space. Therefore, numerical simulation study was conducted to investigate the three-dimensional flow characteristics of the released fluid in the first span space under the double-end fracture accident of the main water supply pipe of the steam generator in the AP1000 nuclear power plant. The three-dimensional model of the first span was established using the ANSYS software. Based on the two-phase flow VOF model, the flow characteristics and water level variations in the first span were calculated in the case of a coolant ejection accident. The calculation results show that the broken water flows through the 5.334 m layer after entering the first span space from the entrance. Most of the discharged water flows out through the reserved openings set in this layer, and part of the water accumulates in this layer. However, due to the setting of the water retaining edge, the floodwater did not flow to the 0 m layer and the -3.8 m layer. As the feed water pump jumps due to the coolant spray, the water level in the first span space will gradually drop, which will not cause important equipment drowning. The calculation results provide a numerical reference for the main flood discharge channels and key equipment layout design and optimization of nuclear power plants.
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