共查询到19条相似文献,搜索用时 93 毫秒
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法国CPY核电厂的双重低温超压保护,即在一回路满水的冷停堆工况下,降低稳压器先导式安全阀的开启/关闭压力整定值,在余热排出系统(RRA)正常运行时由RRA安全阀提供低温超压保护,在RRA因破口或误操作隔离时,则由降低了开启/关闭压力整定值的稳压器安全阀提供低温超压保护。低温超压的瞬态模拟和应力分析的结果显示降低稳压器安全阀的开启/关闭压力整定值能够在低温冷停堆状态下为反应堆冷却剂系统(RCP)提供有效的超压保护,避免反应堆压力容器出现脆性断裂,确保一回路压力边界的结构完整性。 相似文献
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对百万千瓦级核电厂的停堆运行事故风险进行内部事件1级概率安全评价(PSA),并根据不同的停堆进程分别建立停堆PSA模型,分析经历LOI-RRA水位对电厂风险水平构成的影响。分析结果表明停堆工况下的电厂风险不可忽视,在冷停堆工况下经历LOI-RRA水位导致堆芯损坏频率明显增加。 相似文献
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对百万千瓦级压水堆核电站燃料棒的结构特点、技术特性进行了描述,并对该类型燃料棒的主要制造工艺包括电子束焊接、压力电阻焊接、燃料芯块装管等进行论述和分析,最后对该元件的国产化提出意见和建议。 相似文献
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硼稀释事故可在电厂所有运行模式下发生,是对核电厂的安全造成威胁的主要事故之一。本文概要地叙述了压水堆硼稀释事故的原因、后果和在设计中的预防及改进措施。 相似文献
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压水堆核电厂燃料元件破损诊断方法 总被引:9,自引:4,他引:5
在核电厂运行管理中, 如果在停堆前知道燃料棒的性能和状态,采用合适的燃料检测管理策略,可减少反应堆的停运时间.本文以燃料元件破损后裂变产物向冷却剂释放的理论为基础,开发了一种通过分析反应堆冷却剂中裂变产物放射性活度估算破损燃料元件的数量、破损尺寸和位置的方法.用大亚湾核电站1号机组第2循环的运行跟踪数据对U1C2燃料组件进行了破损诊断.结果表明,诊断结果与停堆后的实测结果基本吻合. 相似文献
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对百万千瓦参考核电站长燃耗堆芯采用的可燃毒物含量与堆芯燃料管理主要结果进行分析研究。该研究采用先进的燃料管理程序系统,对不同可燃毒物含量和不同可燃毒物棒根数据的进行了计算,给出了组件无了增殖因子随燃耗的变化关系,据此对参考堆芯采用相同的装载进行了4种方案燃料管理计算。 相似文献
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内部水淹是威胁核电厂安全的风险源之一。根据国内外核电厂内部水淹防护设计的标准及实践,归纳提炼出具有工程参考意义的内部水淹危害性分析方法(源方法和设备分析方法),并以国内百万千瓦级压水堆核电厂特定房间为研究对象,采用源方法进行水淹危害性分析。分析结果表明该房间内的水淹对电厂安全不构成威胁,验证了内部水淹危害性分析方法的合理性和有效性。 相似文献
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K. Zhang X.W. Cao J. Deng Z. Wang L.C. Guo D.Q. Guo J.T. Yuan 《Nuclear Engineering and Design》2008,238(7):1720-1727
Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization. 相似文献
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Sung-Min Cho Seung-Jong Oh Aya Diab 《Journal of Nuclear Science and Technology》2018,55(10):1151-1162
This study focuses on the in-vessel phase of severe accident management (SAM) strategy for a hypothetical 1000 MWe pressurized water reactor (PWR). To examine the effectiveness of SAM strategy, it is necessary to identify and assess epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is reactor coolant system (RCS) depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, sensitivity analyses are carried out to assess the impact of the cutoff porosity related to the flow area of core node (EPSCUT), the critical temperature (TCLMAX) and the minimum fraction of oxidized Zr (FZORUP) for cladding rupture, and the flag to divert gas flows in the core to the bypass channel (FGBYPA) on the core melting and relocation process. In this study, the effect of injection time on the integrity of RPV has been examined based on the quantification of the scenario and phenomenological uncertainties. 相似文献
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根据现有的设计资料,使用一体化严重事故分析程序MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10kg/s)、中流量(50kg/s)和大流量(200kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。 相似文献
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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
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介绍了AP1000核电项目采用开顶法和模块化施工对已安装的物项成品保护造成的困难,以及建设实施过程中面临的主要问题,从管理体系和程序体系及采取的技术措施等方面,介绍了如何做好AP1000项目的成品保护工作。 相似文献
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分析了现有压水堆核电站主回路安装设计的特征,并提出相应改进方案,通过对现有技术和改进技术的对比,阐述了反应堆冷却剂系统主回路安装设计改进技术的优点,希望能对提高压水堆核电站主回路的安装及质量控制水平起到促进作用。 相似文献