A historical review of in-vessel melt retention （IVR） is given, which is a severe accident mitigation mea- sure extensively applied in Generation III pressurized water reactors （PWRs）. The idea of IVR actually ori...A historical review of in-vessel melt retention （IVR） is given, which is a severe accident mitigation mea- sure extensively applied in Generation III pressurized water reactors （PWRs）. The idea of IVR actually originated from the back-fitting of the Generation 11 reactor Loviisa WER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse APIO00, the Korean APR1400 as well as Chinese advanced PWR designs HPRIO00 and CAP1400. The most influential phe- nomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel （RPV）. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contrib- ute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.展开更多
In order to evaluate the effects of the FHs （flow holes） on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this...In order to evaluate the effects of the FHs （flow holes） on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this model, it was mainly clear that in the 40% rated operational conditions, the shape of the FHs on the inner barrel did not change largely to the upper plenum thermal-hydraulics. The effect of the FHs on the honeycomb structure in the upper structure was also investigated in these calculations. The results indicated that the height of thermal stratification interface became lower than that evaluated from the test data.展开更多
The fluoride salt-cooled high-temperature reactor(FHR) uses molten FLi Be salt as the coolant, which introduces a corrosive effect on the alloy-N structure material. Fission neutrons activate the corroded alloy-N,alon...The fluoride salt-cooled high-temperature reactor(FHR) uses molten FLi Be salt as the coolant, which introduces a corrosive effect on the alloy-N structure material. Fission neutrons activate the corroded alloy-N,along with alloy-N structures inside the reactor vessel. The activation products of the alloy-N have a big impact on radiation protection during operation, maintenance, and decommissioning of the reactor. We have constructed a SCALE 6.1 model for the core of a typical 10 MW th FHR and analyzed the activity of each constituent of the irradiated alloy-N. The results show that the activity is predominantly due to short-lived^(28) Al,^(60m) Co,^(56) Mn,^(51)Ti, and ^(52) V, as well as long-lived ^(60) Co,^(51)Cr,^(55)Fe,^(59)Fe, and ^(54) Mn.Furthermore, because of their relatively long half-life and high-energy c-rays emissions,^(60) Co and ^(54)Mn are the major contributors to the radiation source terms introduced by alloy-N activation. The yield of ^(60)Co and ^(54)Mn per unit mass of alloy-N under the current core design is 5.58*10~5 and 1.55 * 10~3 Bq MWd^(-1)g^(-1), respectively.The results of this paper, combined with future corrosion studies, may provide a basis for evaluating long-term radiation source terms of the primary loop salt and components.展开更多
铅冷快堆是第四代核能系统推荐堆型之一,世界上多个铅冷快堆采用非能动余热排出系统。非能动系统中作为驱动的自然力与阻力在数量级上接近,由周边环境、材料参数的变化引起的波动不可忽略,因此需要研究非能动系统可靠性。改进了常用的...铅冷快堆是第四代核能系统推荐堆型之一,世界上多个铅冷快堆采用非能动余热排出系统。非能动系统中作为驱动的自然力与阻力在数量级上接近,由周边环境、材料参数的变化引起的波动不可忽略,因此需要研究非能动系统可靠性。改进了常用的响应面分析法,并应用于中国铅基研究堆反应堆容器空气冷却系统(Reactor Vessel Air Cooling System,RVACS)中。分析中使用流体计算软件Fluent模拟中国铅基研究堆RVACS系统的余热排出过程,研究了输入参数的不确定性对系统可靠性及反应堆安全产生的影响。在大量模拟数据的基础上结合神经网络法建立了输入参数不确定性和结果不确定性之间的映射关系,并以此分析RVACS非能动失效概率。分析结果表明在全厂断电的情况下,RVACS四组并联排热管中的两组也能够可靠地导出反应堆余热。展开更多
文摘A historical review of in-vessel melt retention （IVR） is given, which is a severe accident mitigation mea- sure extensively applied in Generation III pressurized water reactors （PWRs）. The idea of IVR actually originated from the back-fitting of the Generation 11 reactor Loviisa WER-440 in order to cope with the core-melt risk. It was then employed in the new deigns such as Westinghouse APIO00, the Korean APR1400 as well as Chinese advanced PWR designs HPRIO00 and CAP1400. The most influential phe- nomena on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the vessel by the molten core, and the external cooling of the reactor pressure vessel （RPV）. For in-vessel melt evolution, past focus has only been placed on the melt pool convection in the lower plenum of the RPV; however, through our review and analysis, we believe that other in-vessel phenomena, including core degradation and relocation, debris formation, and coolability and melt pool formation, may all contrib- ute to the final state of the melt pool and its thermal loads on the lower head. By looking into previous research on relevant topics, we aim to identify the missing pieces in the picture. Based on the state of the art, we conclude by proposing future research needs.
文摘In order to evaluate the effects of the FHs （flow holes） on the inner barrel, which were installed in the upper plenum of the Monju reactor vessel, a high resolution meshes around the FHs was constructed. Using this model, it was mainly clear that in the 40% rated operational conditions, the shape of the FHs on the inner barrel did not change largely to the upper plenum thermal-hydraulics. The effect of the FHs on the honeycomb structure in the upper structure was also investigated in these calculations. The results indicated that the height of thermal stratification interface became lower than that evaluated from the test data.
基金Supported by the ‘‘Strategic Priority Research Program’’ of the Chinese Academy of Sciences(Grant No.XDA02050100)
文摘The fluoride salt-cooled high-temperature reactor(FHR) uses molten FLi Be salt as the coolant, which introduces a corrosive effect on the alloy-N structure material. Fission neutrons activate the corroded alloy-N,along with alloy-N structures inside the reactor vessel. The activation products of the alloy-N have a big impact on radiation protection during operation, maintenance, and decommissioning of the reactor. We have constructed a SCALE 6.1 model for the core of a typical 10 MW th FHR and analyzed the activity of each constituent of the irradiated alloy-N. The results show that the activity is predominantly due to short-lived^(28) Al,^(60m) Co,^(56) Mn,^(51)Ti, and ^(52) V, as well as long-lived ^(60) Co,^(51)Cr,^(55)Fe,^(59)Fe, and ^(54) Mn.Furthermore, because of their relatively long half-life and high-energy c-rays emissions,^(60) Co and ^(54)Mn are the major contributors to the radiation source terms introduced by alloy-N activation. The yield of ^(60)Co and ^(54)Mn per unit mass of alloy-N under the current core design is 5.58*10~5 and 1.55 * 10~3 Bq MWd^(-1)g^(-1), respectively.The results of this paper, combined with future corrosion studies, may provide a basis for evaluating long-term radiation source terms of the primary loop salt and components.
文摘铅冷快堆是第四代核能系统推荐堆型之一,世界上多个铅冷快堆采用非能动余热排出系统。非能动系统中作为驱动的自然力与阻力在数量级上接近,由周边环境、材料参数的变化引起的波动不可忽略,因此需要研究非能动系统可靠性。改进了常用的响应面分析法,并应用于中国铅基研究堆反应堆容器空气冷却系统(Reactor Vessel Air Cooling System,RVACS)中。分析中使用流体计算软件Fluent模拟中国铅基研究堆RVACS系统的余热排出过程,研究了输入参数的不确定性对系统可靠性及反应堆安全产生的影响。在大量模拟数据的基础上结合神经网络法建立了输入参数不确定性和结果不确定性之间的映射关系,并以此分析RVACS非能动失效概率。分析结果表明在全厂断电的情况下,RVACS四组并联排热管中的两组也能够可靠地导出反应堆余热。
