本文的研究是以“九五”国家科技攻关专题“快堆主钠池堆芯抗震性能的安全评价方法研究”中的抗屈曲研究子课题为背景展开的。所用模型简化自中国实验快堆钠池主容器,是由薄壁圆柱壳和多个加劲肋结合而成的组合圆柱壳体。本文分别采用...本文的研究是以“九五”国家科技攻关专题“快堆主钠池堆芯抗震性能的安全评价方法研究”中的抗屈曲研究子课题为背景展开的。所用模型简化自中国实验快堆钠池主容器,是由薄壁圆柱壳和多个加劲肋结合而成的组合圆柱壳体。本文分别采用大型有限元程序ANSYS 5.4和ALGOR FEAS(SUPER SAP 93),对该壳体进行了常温时水平、轴向荷载共同作用下静态屈曲的计算,同时考虑了诸如塑性、边界条件及初始缺陷等因素的影响。并进行了相关实验研究,最后将有限元计算结果与实验所获得的静屈曲荷载进行比较,结果吻合较好。展开更多
A water leakage on the surface of heat transfer tube in a steam generator of sodium-cooled fast reactor causes SWR （sodium-water reaction）. The SWR damages the leak surface and gives rise to the leak enlargement. Mo...A water leakage on the surface of heat transfer tube in a steam generator of sodium-cooled fast reactor causes SWR （sodium-water reaction）. The SWR damages the leak surface and gives rise to the leak enlargement. Most of initial leakage starts from micro leak （less than 0.5 g/s）. However, the leak rate increases more than two orders of magnitude and the resultant leak damages surrounding heat transfer tubes and it brings secondary failure of the heat transfer tube. Evaluation of the leak enlargement is necessary to assess the leak rate increase, so that evaluate the possibility of secondary failure. In this study, a simulant experiment, which uses neutralization reaction, is proposed to reproduce the leak enlargement. To examine the feasibility of the experiment, numerical simulations are carried out. From the result, a funnel-shaped nozzle enlargement is observed and the shape similar to the shape of the enlarged nozzle from the SWAT （sodium-water reaction test loop） experiment.展开更多
The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, si...The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER （Sn, implicit, multifield, multicomponent, Eulerian recriticality） was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.展开更多
Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. ...Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. A fast reactor system is one of the most promising options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive wastes. Based on the experiences gained during the development of the conceptual designs for KALIMER （Korea advanced liquid metal reactor）, the KAERI （Korea Atomic Energy Research Institute） is currently developing advanced SFR （sodium cooled fast reactor） design concepts that can better meet the Gen IV （Generation IV） technology goals. The long-term advanced SFR development plan will be carried out toward the construction of an advanced SFR demonstration plant by 2028. Advanced concept design studies and the development of the advanced SFR technologies necessary for its commercialization and basic key technologies carried out by KAERI are included in this paper.展开更多
Core damage accident scenarios are identified for the metal-fueled, sodium-cooled fast reactor （SFR）, KALIMER-600, which is under development at KAERI. A level 1 probabilistic safety assessment （PSA） model is deve...Core damage accident scenarios are identified for the metal-fueled, sodium-cooled fast reactor （SFR）, KALIMER-600, which is under development at KAERI. A level 1 probabilistic safety assessment （PSA） model is developed using the identified accident scenarios and the system fault tree models for the safety systems which are needed to mitigate the accidents. Using the preliminary level 1 PSA models, core damage frequency is estimated for the metal fueled KALIMER-600 conceptual design. Sensitivity studies for various design alternatives of safety systems are performed to find out optimal configurations in point of view of risk minimization.展开更多
文摘本文的研究是以“九五”国家科技攻关专题“快堆主钠池堆芯抗震性能的安全评价方法研究”中的抗屈曲研究子课题为背景展开的。所用模型简化自中国实验快堆钠池主容器,是由薄壁圆柱壳和多个加劲肋结合而成的组合圆柱壳体。本文分别采用大型有限元程序ANSYS 5.4和ALGOR FEAS(SUPER SAP 93),对该壳体进行了常温时水平、轴向荷载共同作用下静态屈曲的计算,同时考虑了诸如塑性、边界条件及初始缺陷等因素的影响。并进行了相关实验研究,最后将有限元计算结果与实验所获得的静屈曲荷载进行比较,结果吻合较好。
文摘A water leakage on the surface of heat transfer tube in a steam generator of sodium-cooled fast reactor causes SWR （sodium-water reaction）. The SWR damages the leak surface and gives rise to the leak enlargement. Most of initial leakage starts from micro leak （less than 0.5 g/s）. However, the leak rate increases more than two orders of magnitude and the resultant leak damages surrounding heat transfer tubes and it brings secondary failure of the heat transfer tube. Evaluation of the leak enlargement is necessary to assess the leak rate increase, so that evaluate the possibility of secondary failure. In this study, a simulant experiment, which uses neutralization reaction, is proposed to reproduce the leak enlargement. To examine the feasibility of the experiment, numerical simulations are carried out. From the result, a funnel-shaped nozzle enlargement is observed and the shape similar to the shape of the enlarged nozzle from the SWAT （sodium-water reaction test loop） experiment.
文摘The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT （control-rod guide tube） during CDAs （core-disruptive accidents） of SFRs （sodium-cooled fast reactors）, since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER （Sn, implicit, multifield, multicomponent, Eulerian recriticality） was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.
文摘Korea imports about 97% of its energy resources as its available energy resources are extremely limited. Thus, the role of nuclear power in electricity generation is expected to become more important in future years. A fast reactor system is one of the most promising options for electricity generation with an efficient utilization of uranium resources and a reduction of radioactive wastes. Based on the experiences gained during the development of the conceptual designs for KALIMER （Korea advanced liquid metal reactor）, the KAERI （Korea Atomic Energy Research Institute） is currently developing advanced SFR （sodium cooled fast reactor） design concepts that can better meet the Gen IV （Generation IV） technology goals. The long-term advanced SFR development plan will be carried out toward the construction of an advanced SFR demonstration plant by 2028. Advanced concept design studies and the development of the advanced SFR technologies necessary for its commercialization and basic key technologies carried out by KAERI are included in this paper.
文摘Core damage accident scenarios are identified for the metal-fueled, sodium-cooled fast reactor （SFR）, KALIMER-600, which is under development at KAERI. A level 1 probabilistic safety assessment （PSA） model is developed using the identified accident scenarios and the system fault tree models for the safety systems which are needed to mitigate the accidents. Using the preliminary level 1 PSA models, core damage frequency is estimated for the metal fueled KALIMER-600 conceptual design. Sensitivity studies for various design alternatives of safety systems are performed to find out optimal configurations in point of view of risk minimization.
