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全厂断电(SBO)可能发展成为堆芯熔化、安全壳超压失效的严重事故。本文首先研究全厂断电事故的必要性以及在辅助给水系统不可用情况下的全厂断电事故的进程,随后定性的分析了事故进程在主泵轴封泄漏和对一回路实施减压缓解措施的影响下所具有的不同的发展情况。最后以秦山核电厂为例对其在提高应对全厂断电事故的能力和改进缓解事故后果的措施方面提出了建议。 相似文献
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在发生全厂断电的情况下,冷管段出现破口将会进一步加快事故进程。利用一体化严重事故分析程序MAAP4对百万千瓦级核电站全厂断电叠加冷管段破口进行计算分析,得到该事故时间序列和关键热工水力参数随时间的变化趋势。对于重要参数(一回路压力,堆芯液位,时间序列等)的分析:随着中小破口当量直径由4 cm增至5 cm,堆芯裸露时间分和失效时间提前分别约1 000 s和3000 s;中破口当量直径由5 cm增至7 cm,堆芯裸露时间和失效时间提前分别约1 400 s和6 457 s;而大破口事故当量直径由20 cm增至21 cm,堆芯裸露时间和失效时间分别仅提前约20 s和230 s。相关数据及其分析可为严重事故的缓解措施提供相关理论依据。 相似文献
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辅助给水系统对缓解全厂断电事故能力研究 总被引:1,自引:1,他引:0
以CPR1000核电站为研究对象,采用RELAP5/MOD3.2轻水堆瞬态分析程序,对系统进行合理简化并建模,模拟系统在全厂断电事故下的瞬态响应过程,研究全厂断电事故发生后辅助给水(AFW)的投入对缓解全厂断电事故的能力。计算结果表明:断电事故发生后,主给水丧失导致一回路压力和冷却剂平均温度在断电后6s达到峰值;辅助给水投入约200s后,一回路因热阱丧失而引起的温度和压力升高能有效地得到缓解,为交流电源的恢复及余热排出系统的投入赢得了更多的时间。 相似文献
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大亚湾核电站全厂断电事故及第5台应急柴油机的概率安全评价 总被引:5,自引:2,他引:3
对大亚湾核电站全厂断电事故(SBO)及第5台应急柴油机改进项目进行了概率安全评价(PSA),给出了电源不可恢复因子的计算方法,并对第5台应急柴油机的接入时间进行了敏感性分析。研究结果表明,全厂断电引起的堆芯损坏频率(CDF)较大,增加第5台柴油机对降低堆芯损伤风险有明显的好处.而该台柴油机接入时间的长短对降低堆芯损坏频率有较大影响。 相似文献
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采用严重事故最佳估算程序RELAP5/SCDAPSIM/MOD3.2,建立美国Surry-2核电站的详细计算模型,对完全丧失给水(TLFW)引发的堆芯熔化事故进行研究分析。为准确预测压力容器内堆芯熔化的进程,为二级概率安全评价提供可信的初始条件,计算中考虑了一回路压力边界的蠕变破裂失效,并评价了人为干预对堆芯熔化进程及事故后果的影响。计算结果表明,由完全丧失给水引发的压水堆核电站严重事故不会出现人们担心的高压熔堆;反应堆压力容器下封头的失效位置不是在其底部,而是在其侧面;通过打开稳压器释放阀对一回路实施主动卸压能够大大推迟事故的进程。 相似文献
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Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima. 相似文献
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Using the MELCOR code, we simulated and analyzed a severe accident at a Chinese pressurized reactor 1000-MW (CPR1000) power plant caused by station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV). The CPR1000 response and results for three different scenarios were analyzed: (i) seal leakage and an auxiliary feed water (AFW) supply; (ii) no seal leakage or AFW supply; and (iii) seal leakage but no AFW supply. The results for the three scenarios are compared with those for a simple SBO accident. According to our calculations, the SG SRV stuck in the open position would greatly accelerate the sequence for a severe accident. For an SBO accident with the SRV stuck open without seal leakage or an AFW supply, the pressure vessel would fail at 9576 s and the containment system would fail at 124,000 s. If AFW is supplied, pressure vessel failure would be delayed nearly 30000 s and containment failure would delay at least 50000 s. When seal leakage exists, pressure vessel failure is delayed about 50 s and containment failure time would delay about 30000 s. The results will be useful in gaining an insight into the detailed processes involved and establishing management guidelines for a CPR1000 severe accident. 相似文献
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堆芯熔化严重事故下保证反应堆压力容器(RPV)完整性非常重要,高温蠕变失效是堆芯熔化严重事故下反应堆压力容器的主要失效模式。在进行严重事故堆芯熔化物堆内包容(IVR)下RPV结构完整性分析中,RPV内外壁和沿高度方向的温度分布以及剩余壁厚是结构分析的重要输入。本文采用CFD分析方法对RPV堆内熔融物、RPV壁以及外部气液两相流动换热进行热-固-流耦合分析,获得耦合情况下的温度场、流场、各相份额分布以及RPV的剩余壁厚,为RPV在严重事故IVR下的结构完整性分析提供依据。 相似文献
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压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列:热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应(MCCI),安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。 相似文献
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本文利用MELCOR1.8.5程序建立了典型的M310核电站的严重事故模型,基于该模型设计了多种非能动的缓解措施,针对由全厂断电诱发的严重事故,模拟研究了这些非能动安全措施的缓解效果。研究结果表明:在全厂断电事故下,堆芯补水箱系统、堆腔注水系统、非能动余热排出系统均能有效地投入使用,并显著地延缓事故的发展,将核电站稳定在一个安全的状态,为人工干预赢得更多时间。 相似文献
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Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible failure mode of the reactor pressure vessel (RPV) and its failure time have to be investigated for a determination of the load conditions for subsequent containment analyses. Worldwide several experiments have been performed in this field accompanied with material properties evaluation, theoretical, and numerical work. 相似文献