The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, ...The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of the more formidable and dangerous operation environments of them. This paper presents results of marine black out accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4 code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation.展开更多
为了更好地研究事故条件下非能动安全壳热量导出系统作用下安全壳内的热工水力行为,中国核电工程有限公司搭建安全壳综合性能实验装置(PlAtform for iNteGral TH behaviour of containment,PANGU)并开展了3种事故序列大破口事故(堆芯未...为了更好地研究事故条件下非能动安全壳热量导出系统作用下安全壳内的热工水力行为,中国核电工程有限公司搭建安全壳综合性能实验装置(PlAtform for iNteGral TH behaviour of containment,PANGU)并开展了3种事故序列大破口事故(堆芯未熔)、大破口事故(堆芯熔化)和全厂断电事故下的实验研究。采用GOTHIC程序建立安全壳综合性能实验装置数值计算模型,并针对已开展的3个实验进行数值计算研究,得出结论如下:对于3个事故序列,程序计算的穹顶区域水蒸气浓度与实验值趋势上保持一致,特别是长期阶段水蒸气浓度实验值与计算值符合良好;计算模型所计算的安全壳内温度压力无论是峰值还是长期值均与实验值保持在较小的误差范围内;简化后的PCS模型计算的PCS功率略低于实验测量的PCS功率,72 h内计算的PCS总排热量与实验测量值相当。本文研究结果可为“华龙一号”PCS系统计算分析提供理论支持。展开更多
The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal sh...The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal shock （PTS） events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States （US） rules： Operators may act after 10 min （Japanese rules） and 30 min （the US rules） after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.展开更多
文摘The design of an integral pressurized water reactor (IPWR) focuses on enhancing the safety and reliability of the reactor by incorporating a number of inherent safety features and engineered safety features. However, the characteristics of passive safety systems for the marine reactors are quiet different from those for the land nuclear power plant because of the more formidable and dangerous operation environments of them. This paper presents results of marine black out accident analyses. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the RELAP5/MOD3.4 code and a sensitivity analysis has been performed to evaluate a passive cooldown capability. The results of the accident analyses show that the reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation.
文摘为了更好地研究事故条件下非能动安全壳热量导出系统作用下安全壳内的热工水力行为,中国核电工程有限公司搭建安全壳综合性能实验装置(PlAtform for iNteGral TH behaviour of containment,PANGU)并开展了3种事故序列大破口事故(堆芯未熔)、大破口事故(堆芯熔化)和全厂断电事故下的实验研究。采用GOTHIC程序建立安全壳综合性能实验装置数值计算模型,并针对已开展的3个实验进行数值计算研究,得出结论如下:对于3个事故序列,程序计算的穹顶区域水蒸气浓度与实验值趋势上保持一致,特别是长期阶段水蒸气浓度实验值与计算值符合良好;计算模型所计算的安全壳内温度压力无论是峰值还是长期值均与实验值保持在较小的误差范围内;简化后的PCS模型计算的PCS功率略低于实验测量的PCS功率,72 h内计算的PCS总排热量与实验测量值相当。本文研究结果可为“华龙一号”PCS系统计算分析提供理论支持。
文摘The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal shock （PTS） events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States （US） rules： Operators may act after 10 min （Japanese rules） and 30 min （the US rules） after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.
