共查询到18条相似文献,搜索用时 281 毫秒
1.
船用堆破口叠加全船断电事故进程及后果研究 总被引:2,自引:2,他引:0
采用MELCOR程序,对船用堆破口叠加全船断电事故进行建模计算,并对事故进程和源项释放进行了研究。计算结果表明:若应急电源无法投入,最终将导致压力容器下封头失效和舱底失效;所研究事故的惰性气体、碘释放量均在80%以上,且释放的I主要以CsI形式存在,滞留量大,气载量小。事故进展快慢取决于破口当量尺寸,但氢气的产量与堆芯温度、堆芯残余水量相关,与破口当量尺寸无直接关系,堆舱内发生氢爆可能性不大。本文计算结果可为应急抢修和应急决策提供技术支持。 相似文献
2.
以严重事故分析程序MELCOR为计算工具,研究了某型船用堆发生蒸汽发生器传热管破损叠加全船断电事故,针对传热管破损所导致的放射性物质向其他舱室的泄漏,着重分析了惰性气体和CsI的释放、迁移、滞留特点及其在舱室内的分布。计算结果表明:二回路蒸汽管道会发生超压失效,氢燃导致堆舱邻舱的超压失效。至计算结束,约占累积总量99.61%的Xe和49.96%的CsI从堆芯释放出来。舱室Ⅰ和Ⅱ内Xe的分布份额分别为38%和18%,CsI的分布份额分别为22.2%和2.7%,CsI主要存在于舱底水池中,且泄漏至舱室Ⅱ的份额微少。本文分析结果可为进一步的源项剂量分析及船内外应急提供依据。 相似文献
3.
《核电子学与探测技术》2017,(12)
严重事故情况下大量放射性裂变产物释放到环境中,直接对人体造成危害。本文采用MELCOR程序,研究了DVI管小破口始发严重事故下,Cs I在一回路系统和安全壳中的存在形态,以及Cs I和惰性气体向安全壳、环境的迁移和释放行为。结果表明:Cs I主要以气溶胶形式释放出来,滞留在一回路系统中的Cs I几乎全部沉积在热构件上,约70%的Cs I以气溶胶的形式迁移到安全壳中,并不断沉积在安全壳热构件或溶于水中,极少量释放到环境中。事故后绝大部分惰性气体迁移到安全壳中,只有少部分滞留在反应堆其他系统中,在安全壳正常泄漏率下,释放到环境的惰性气体质量很少,仅为0.11%。 相似文献
4.
5.
文章在对轻水堆核电站先进堆型AP1000失水事故(LOCA)的事故进程分析的基础上,明确了失水事故堆芯释放源项的核素类型,再基于《AP1000设计手册》中提供的基础设计数据,利用ORIGEN2编程对关注的核素进行计算,求取各核素在0~8 h内放射性活度随时间的变化。并将计算结果与设计值进行对比分析,从结果来看,大部分核素的计算值与设计值数量级基本吻合,部分核素的计算值与设计值之间存在1~2个数量级的差异,这是因为在源项选择中忽略了部分核素,此外,选取的堆芯放射性核素的积存量为保守的基准设计值。核电站应当加强对碱金属、惰性气体和碘的关注。在事故前期,碱金属138Cs约占总放射性的85.6%;事故后期,则是惰性气体133Xe占比最大,约为53.1%。 相似文献
6.
7.
8.
9.
10.
事故期间安全壳内的辐射水平是堆芯损伤评价和进行防护决策的重要依据,计算不同堆芯状况下安全壳内辐射监测仪表示值是应用该方法的前提条件.文章比较了正常冷却剂释放、间隙释放和堆芯熔化状况下不同核素对安全壳内辐射监测仪表示值的相对贡献.在安全壳内无喷淋情况下,安全壳内辐射监测仪表示值主要来自碘和惰性气体;安全壳内有喷淋情况下的辐射监测仪表示值主要来自于惰性气体. 相似文献
11.
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
12.
13.
14.
福岛核电厂3号机组严重事故模拟分析 总被引:1,自引:1,他引:0
本文应用MELCOR程序,通过建立全厂详细的模型,对福岛第一核电厂3号机组在地震发生后3 d内的严重事故进程进行了模拟分析并与电厂实测数据进行了比较,再现了从事故开始到堆芯失效坍塌直至氢气爆炸在内的主要严重事故现象。基于文中假设的模拟计算得到的趋势与电厂现有实测数据较为一致,结果表明:地震发生后约36 h反应堆水位降至堆芯活性区顶部。操纵员未能及时成功对安全壳和反应堆进行快速卸压,以在堆芯底部出现裸露前向反应堆补充冷却水,使得堆芯出现严重的锆水反应,大部分燃料包壳已破损而导致易挥发的放射性裂变产物的释放;但此时堆芯整体依然维持可冷却几何形状;在消防水泵向反应堆注入冷却水期间,由于冷却注入流量出现中断,导致堆芯进一步熔毁坍塌;碎片迁移至下腔室后,进一步的锆水反应(金属 水反应)新增的氢气泄漏并积聚在反应堆厂房上部,引发了安全壳厂房的爆炸;72 h内,堆芯内约50%的锆合金发生了氧化,压力容器下封头未发生失效。 相似文献
15.
16.
以先进压水堆核电厂为对象,开展了适用于应急设施可居留性评价的严重事故源项分析方案研究,覆盖了堆芯释放、安全壳内自然去除、放射性物质向环境释放途径等。结合非能动安全壳冷却系统的特征,重点研究了安全壳可能的失效行为,论证了安全壳在事故后24h和72h失效工况的辐射影响。结果表明:两种工况放射性释放水平均达到了INES(国际核事件分级)第6级的水平,属于比较严重的核事故;133 Xe、131I为主导核素组的主导核素,所释放的133 Xe介于WASH-1400中PWR2~PWR4之间的水平,131I介于PWR5~PWR6之间水平。同时,以国内某沿海厂址为例,评价了两种工况下应急指挥中心(EOF)工作人员的有效剂量,均可满足100mSv的剂量限值要求。 相似文献
17.
Containment depressurization has been implemented for many nuclear power plants (NPPs) to mitigate the risk of containment overpressurization induced by steam and gases released in LOCA accidents or generated in molten core concrete interaction (MCCI) during severe accidents. Two accident sequences of large break loss of coolant accident (LB-LOCA) and station blackout (SBO) are selected to evaluate the effectiveness of the containment venting strategy for a Chinese 1000 MWe NPP, including the containment pressure behaviors, which are analyzed with the integral safety analyses code for the selected sequences. Different open/close pressures for the venting system are also investigated to evaluate CsI mass fraction released to the environment for different cases with filtered venting or without filtered venting. The analytical results show that when the containment sprays can't be initiated, the depressurization strategy by using the Containment Filtered Venting System (CFVS) can prevent the containment failure and reduce the amount of CsI released to the environment, and if CFVS is closed at higher pressure, the operation interval is smaller and the radioactive released to the environment is less, and if CFVS open pressure is increased, the radioactive released to the environment can be delayed. Considering the risk of high pressure core melt sequence, RCS depressurization makes the CFVS to be initiated 7 h earlier than the base case to initiate the containment venting due to more coolant flowing into the containment. 相似文献
18.
William J. Galyean Christiana H. Lui Thomas D. Brown Douglas A. Brownson Julie J. Gregory 《Nuclear Engineering and Design》1999,194(1):97
The US Nuclear Regulatory Commission’s Accident Sequence Precursor (ASP) program currently uses simple level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Not all accident sequences leading to core damage will result in the same radiological consequences. Therefore, it is necessary to develop simple level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. This level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Two versions of these prototype level 2/3 models have been developed. Simple level 2/3 models have been linked to the simple level 1 models to provide risk perspectives for operational events. Additionally, very complex and detailed models have been developed that take provide much greater flexibility in accommodating a much wider range of level 1 core damage accident sequences. These detailed models also make possible assessments that are integrated with the level 1 model, on the importance of different severe accident phenomena and containment performance characteristics. This paper describes the development and capabilities of these level 2/3 ASP models, and the linkage to the level 1 models. 相似文献