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失水事故是核电站设计的基准事故之一,是压水堆事故分析关注的重点.本文概括介绍了秦山核电二期工程的失水事故分析及分析计算所使用的计算程序;简要地描述了MEFRA-1等计算程序的特点.重点介绍了大破口失水事故分析,给出了分析计算的主要假设条件和分析计算结果.分析计算表明,大破口失水事故工况下,燃料元件最大峰值包壳表面温度为1092.56℃,秦山核电二期工程的安全注射系统能保证该核电站在发生失水事故时的安全. 相似文献
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基于小型压水堆特有的截短型燃料组件,建立乏燃料贮存水池几何模型,分析正常贮存及事故工况下的临界安全。选取合理的保守假设,建立适当的计算模型,分别计算了一区和二区正常贮存工况、地震事故工况、组件跌落事故工况、新组件误插入事故工况的反应性。计算得到事故工况下有效增值因子最大值为0.932 83。小型压水堆乏燃料贮存水池临界安全分析中,正常工况及事故工况下计算结果均小于0.95。该设计模型可确保燃料堆内贮存区域处于次临界状态,且安全可控。 相似文献
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秦山核电厂调试后失水事故计算分析中采用了高压安注系统和安注箱试验的测量结果,重新分析了大、小破口失水事故。为使分析计算与FSAR有一个可比性,模拟计算采用的初始条件、计算模型及分析程序都与FSAR相同。计算分析的结果进一步确认了秦山核电厂大、小破口失水事故后的安全性,并为FSAR中大、小破口失水事故分析提供了修改的依据。另外,依据秦山核电厂ECCS设计特点和运行方式,并参照LWR失水事故安全准则,评述了秦山核电厂ECCS的设计能力、可靠性和冗余度。 相似文献
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秦山核电厂调试后失水事故计算分析中采用了高压安注系统和安注箱试验的测量结 果,重新分析了大、小破口失水事故。为使分析计算与FSAR有一个可比性,模拟计算采用的初始条件、计算模型及分析程序都与FSAR相同。计算分析的结果进一步确认了秦山核电厂大、小破口失水事故后的安全性,并为FSAR中大、小破口失水事故分析提供了修改的依据。另外,依据秦山核电厂ECCS设计特点和运行方式,并参照LWR失水事故安全准则,评述了秦山核电厂ECCS的设计能力、可靠性和冗余度。 相似文献
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当压水堆核电厂发生事故后,带有放射性的核素会通过破损处释放到环境中,从而危害核电厂周边环境及相关人员的安全,因此对事故后释放到环境中的放射性源项分析,对于核电厂的辐射防护具有重要意义。本文根据事故发生的频率以及后果严重程度,选取蒸汽发生器传热管破裂事故(Steam Generator Tube Rupture,SGTR)进行分析。事故分为事故前碘尖峰释放和事故并发碘尖峰释放两种事故工况,建立事故后放射性核素迁移和扩散计算模型,同时使用先进压水堆AP1000参数进行计算验证,并重点关注惰性气体和挥发性核素碘在环境中的放射性活度。计算结果显示:使用文中计算模型计算的放射性源项与设计源项比较一致,在两种工况下,惰性气体的释放活度与设计源项吻合较好,但碘的释放活度有明显差别。 相似文献
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介绍了美国核管会用于核与辐射事故后果分析的辐射评价系统(RASCAL)的主要功能和特性,重点分析了RASCAL的源项计算剂量模块、场外监测数据计算剂量模块、气象数据处理模块,以及源项计算模式、大气输运扩散模式和剂量计算模式。最后,将RASCAL应用于我国某核电厂事故应急演习中,评价分析事故情景下的放射性影响,并将其结果通过Google Earth进行三维展示。 相似文献
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《Annals of Nuclear Energy》2002,29(17):2055-2069
The most operator support systems including the training simulator have been developed to assist the operator and they cover from normal operation to emergency operation. For the severe accident, the overall architecture for severe accident management is being developed in some developed countries according to the development of severe accident management guidelines which are the skeleton of severe accident management architecture. In Korea, the severe accident management guideline for KSNP was recently developed and it is expected to be a central axis of logical flow for severe accident management. There are a lot of uncertainties in the severe accident phenomena and scenarios and one of the major issues for developing a operator support system for a severe accident is the reduction of these uncertainties. In this paper, the severe accident management advisory system with training simulator, SAMAT, is developed as all available information for a severe accident are re-organized and provided to the management staff in order to reduce the uncertainties. The developed system includes the graphical display for plant and equipment status, the previous research results by knowledge-base technique, and the expected plant behavior using the severe accident training simulator. The plant model used in this paper is oriented to severe accident phenomena and thus can simulate the plant behavior for a severe accident. Therefore, the developed system may make a central role of the information source for decision-making for a severe accident management, and will be used as the training simulator for severe accident management. 相似文献
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This paper presents a methodology utilizing an accident management strategy in order to determine accident environmental conditions to be used as inputs to equipment survivability assessments. In the case that there is a well-established accident management strategy for a specific nuclear power plant (NPP), an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for the accident management strategy or appropriate actions. For this work, three different tools are introduced; probabilistic safety assessment (PSA) outcomes, major accident management strategy actions, and accident environmental stages (AESs). In order to quantitatively investigate an applicability of accident management strategy on equipment survivability, the accident simulation for most likely scenario in Korean standard nuclear power plants (KSNPs) is performed with the MAAP4 code. The accident management guideline (AMG) actions such as the reactor coolant system (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparison to actions from previous normal accident simulation, especially focusing on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages. This implies that plant-specific AMG actions need to be considered in order to determine accident environmental conditions in equipment survivability assessments. 相似文献
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The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines. 相似文献
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