共查询到16条相似文献,搜索用时 171 毫秒
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本文采用严重事故一体化分析软件MAAP4(Modular Accident Analysis Program)对百万千瓦级压水堆进行分析,选取一回路大破口严重事故进行仿真,获得了该事故工况下核电厂关键参数的瞬态特性,与RELAP5计算结果进行了对比验证。在分析MAAP4模型的基础之上,进一步仿真该电站大破口事故后期进程,截取压力边界内外参数进行评估。分析结果表明:MAAP4在仿真安全壳和氢气分布上,预测事故结果置信度高,其中模拟的安全防护设计能够有效缓解事故进程,满足一般核电厂的安全评估要求,对概率安全评价(PSA)具有一定的参考意义。 相似文献
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模块化小型核反应堆(SMR)与传统大型压水堆在结构上存在很大差异,导致两者的严重事故进程存在较大差异。因此本文结合SMR自身设计特点,建立反应堆严重事故分析模型,对SMR的典型事故瞬态进行模拟计算,并对严重事故进程、热工水力现象和系统安全进行研究。在此基础上提出了SMR自动卸压系统优化改进方案,通过对自动卸压系统各级卸压管线的位置和阀门有效面积进行深入研究,并对相关参数进行敏感性分析,提出符合反应堆自身特点的卸压阀门有效面积的优化设计方案,为小型核反应堆的严重事故预防和缓解提供有效的依据和参考。 相似文献
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基于一体化严重事故分析程序MAAP4.0.3(Modular Accident Analysis Program),本文建立了我国现役典型百万千瓦级压水堆(Pressurized-Water Reactor,PWR)核电机组模型,研究了热管段不同面积破口事故叠加安注失效的工况引起的严重事故过程,探讨了如何在恰当的时机采取有效的缓解措施对事故的进程进行干预。研究结果表明:在破口事故中随着破口面积而增大,压力容器会更早失效导致堆芯裸露;一旦压力容器失效,MCCI(Molten Corium Concrete Interaction)过程中氢气产量则会随着破口面积的增大而增大;在破口事故中尽早投入安全注射系统可以有效地缓解事故的进程,避免压力容器失效,并且安全注射系统越早投入对事故的缓解也就越有利。 相似文献
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严重事故下一回路管道可能会发生蠕变失效,若出现蠕变诱发的蒸汽发生器传热管破裂(SGTR),则会导致安全壳旁路失效;若出现蠕变诱发热段或波动管的失效,则产生的破口将会使一回路迅速卸压。因此,评估严重事故下蠕变诱发反应堆冷却剂系统(RCS)破裂的可能性是开展严重事故分析、特别是二级概率安全分析(PSA)的重要基础。本工作基于蠕变失效模型,考虑传热管的缺陷,建立了评价蠕变诱发RCS破裂的确定论模型。在此基础上,运用拉丁超立方体抽样方法,考虑重要参数的不确定性,开发了严重事故下蠕变诱发RCS破裂的概率评估程序。随后对典型的事故序列进行了蠕变诱发RCS破裂的概率评估。结果表明,对于高压事故序列,存在一定的蠕变诱发SGTR概率,也存在较高的蠕变诱发热段或波动管失效概率。 相似文献
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开展了模块化小堆稳压器波动管双端破口试验研究,获得了非能动安全系统的事故响应特性和一回路系统参数变化。试验研究结果表明,在稳压器波动管双端破口极端工况条件下,中压安注箱能在短时间内提供较大的稳定安注流量,及时补充系统水装量;高压安注系统运行过程比较复杂,安注流量与堆芯补水箱压力平衡管线内介质状态和中压安注系统运行状态密切相关,在1.7 h内呈间歇注入运行状态。在整个事故过程中,堆芯一直处于淹没状态,模块化小堆非能动安全系统能够确保稳压器波动管在双端破口极端工况条件下的堆芯安全。 相似文献
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稳压器是核反应堆进行压力控制和保护的重要设备,冷却剂丧失事故(LOCA)产生的巨大冲击可能造成其关键部位的结构失效。通过多场耦合计算方法,对小破口LOCA下稳压器波动管的流动传热和结构应力、人孔结构的温度分布和密封性能进行了三维瞬态数值模拟,分析了其失效机理。结果表明:高温流体快速流入波动管形成了巨大的瞬时载荷,造成了管道短时间的强烈振动,管道中间部位变形最大,可能破坏管道支撑结构;各部位等效应力快速增大,与主管道的接管部位出现了集中应力现象,较大的应力波动会影响其寿命;人孔结构出现较大的温度分布不均匀性,密封结构下垫片的密封性能变化最大,在100 s前后其内、外侧密封面接触压力都降至设计密封比压值以下,即出现泄漏。本文根据分析结果提出了波动管和人孔结构的改进建议,可为船用核动力装置发生小破口LOCA后的事故缓解提供技术借鉴。 相似文献
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全厂断电引发的严重事故下蒸汽发生器传热管蠕变失效风险研究 总被引:1,自引:1,他引:0
全厂断电引发的严重事故若处置不当,可能发展为长期、高压的严重事故进程,此时堆芯冷却系统中的自然循环在导出部分堆芯余热的同时,也增加了蒸汽发生器(SG)传热管、稳压器波动管以及热管段出现蠕变失效的风险。本文基于两环路设计的秦山二期核电厂设计特点,结合蠕变失效风险模型,对全厂断电引发的严重事故后未能执行“严重事故管理导则中向蒸汽发生器注水(SAG-1)”时SG传热管的蠕变失效风险进行了研究,从而为全厂断电引发的严重事故的负面影响提供量化结果,为技术支持中心(TSC)最终决策提供参考依据。分析结果表明,全厂断电引发的严重事故后16 361 s可能出现蠕变失效;自事故后16 610 s,SG传热管出现蠕变失效的可能性均远低于稳压器波动管与热管段,秦山二期核电厂全厂断电引发的严重事故下因SG传热管蠕变失效而导致安全壳旁通的风险很小。 相似文献
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For the passive AP600 plant, the three stages of ADS (automatic depressurization system) valves are attached to the top of pressurizer. The existence of these valves makes liquid flow into and out of the pressurizer an important part of the dynamics during a small break loss-of-coolant accident. In this paper, counter-current flow limit (CCFL) in the surge line was analyzed. Specifically, CCFL in vertical piping, in slightly inclined horizontal piping, and in horizontal and vertical elbows were compared. The CCFL in the vertical section of the surge line was found to be the most limiting section. That is, the vertical CCFL controls the pressurizer liquid drain rate. This conclusion was tested and verified by comparing the predicted vertical CCFL against the counter-current flow states in the surge line, observed in small break LOCA tests conducted at the AP600 scaled test facility (APEX). 相似文献
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《核技术(英文版)》2016,(4):184-194
Thermal mixing and stratification phenomena may occur during the loss of a coolant accident or main steam line break accident in the containment of a Passive Containment Cooling System, or in the suppression pools in BWR. However, the present study pays insufficient attention to the thermal stratification phenomena in the containment of small modular reactors(SMR). In this paper, an investigation on the mixing and thermal stratification phenomena caused by the plumes or buoyant jets in SMR containments was carried out. The experiments were both conducted under non-adiabatic and adiabatic conditions for a steel containment. In each condition, two key parameters, inlet temperature, and flow rate were tested by controlling variables to identify their influence on the thermal stratification phenomenon. The visualization experiments illustrated the jet mixing and stratification development. The experiment results were compared with the numerical computation and they reached a good agreement. 相似文献
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Nicolas Tauveron 《Nuclear Engineering and Design》2008,238(11):2925-2934
This work concerns the design and safety analysis of a direct cycle gas cooled reactor and is concentrated on surge occurrence, development, consequences, and control in case of a pipe rupture. To describe these phenomena, a specific turbomachine model is used. The dynamic behaviour of the compressor and different anti-surge configurations are tested. The tendencies (role of actuators, sensors, gain, bandwidth, plenum size) are in a good agreement with bibliographical data. The application to transient simulations of gas cooled nuclear reactors (hypothetical 10-in. break accident) shows that deep surge has to be avoided to protect compressor, heat exchangers and other pipes, whenever possible. The effects of using the bypass line are pointed out. Different strategies for instrumentation and control deployment are studied and different levels of efficiency are obtained. The complete avoidance of any surge oscillation would require the use of multiple very fast valves with a dedicated control command. 相似文献