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严陈昌 《核标准计量与质量》2012,(1):2-7
对钠冷快堆的原理和主要技术特征进行了阐述,并简要介绍了国内外钠冷快堆的发展概况及其标准与规范现状,最后在借鉴我国压水堆核电厂标准体系建设规划的指导思想和发展思路的基础上,提出了开展钠冷快堆标准体系预研工作的基本设想以及钠冷快堆标准体系顶层设计的初步考虑. 相似文献
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中国一体化反应堆核电厂创新安全壳设计研究 总被引:1,自引:1,他引:0
中国一体化反应堆核电厂(CIP)是中国核反应堆系统设计技术国家重点实验室正在开发的新一代革新型、完全一体化的压水堆,其电功率约为300 MW.CIP采用堆内一体化布置,反应堆冷却剂系统设备以及控制棒驱动机构全部布置在反应堆压力容器内.这种一体化设计消除了传统的冷却剂回路管道,消除了大LOCA事故,具有更高的安全性.本文介绍了CIP安全壳系统方案选择、安全壳设计、安全壳设计压力的确定以及安全壳结构的计算分析. 相似文献
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钠冷快堆通过采用模块式蒸汽发生器的设计方案以提高核电厂的负荷因子。核电厂运行中若发生丧失蒸汽发生器模块事件,核电厂工况将发生变化,应进行适当的调节,调节的目标工况可通过设计与研究给出。本工作对某典型池式钠冷快堆丧失1个蒸汽发生器模块后的最佳工况进行研究,主要研究内容包括对其主热传输系统进行建模,开展主热参数匹配计算,根据相关运行限值来筛选方案并分析关键参数,最终给出较为合适的运行工况。本工作为钠冷快堆在丧失蒸汽发生器模块后的工况设计提供了重要依据。 相似文献
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对我国核电厂事故后安全壳内氢气浓度测量方面的技术水平和发展现状进行了全面调研,分析了事故后安全壳内氢气浓度测量的要求及关键技术难点,提出了3种相关的测量方案,并比较了方案的优缺点。经过比较分析,基于一种探头型分析装置的直接测量方案能够较为准确地实时反映核电厂安全壳内氢气浓度,其发展趋势是应用于未来的大型先进压水堆核电厂中。 相似文献
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张琨 《核标准计量与质量》2012,(3):16-21
安全壳直接加热(DCH)是压水堆核电厂严重事故中的主要现象之一,可能导致安全壳早期失效、大量放射性释放的严重后果.国际上的核安全管理机构均非常重视该现象并制定了相关的法规要求.文章一方面概述与DCH相关的法规要求,另一方面针对AP1000核电厂发生DCH的事故工况,进行后果分析方法的研究.分析结果表明,AP1000核电厂的DCH不会造成安全壳失效. 相似文献
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压水堆核电厂失水事故后安全壳内产氢量计算研究 总被引:2,自引:0,他引:2
采用ORIGEN2程序对压水堆核电厂失水事故工况下堆芯区和地坑区氢气的产生量进行计算,以合理减少安全壳内可燃气体的控制设计评价的保守性.通过冷却剂的辐照分解产氢以及其他相关计算模型,对600MW(电功率)级压水堆核电厂失水事故工况下的氢气产生量进行计算.计算结果表明原评价结果过于保守,在核电厂失水事故后仍有充分的时间准备投入安全壳内氢气复合器. 相似文献
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三种堆型核电厂经济性评价 总被引:1,自引:0,他引:1
本文分析了核电投资的特点、建立了考虑价格浮动和通货膨胀等因素影响的核电厂建成价和核燃料成本的计算模型。对压水堆、高温气冷堆和快堆三种堆型的经济性进行了研究。结果表明,当高温气冷堆和快堆两种先进堆型实现商用概念设计后,其商业竞争能力可与现有的压水堆相媲美。 相似文献
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依据先进非能动压水堆的严重事故管理导则(SAMG),消防系统中的防火喷淋系统,尽管属于非安全相关的系统,仍可以作为严重事故缓解策略,在以下三个方面起到严重事故缓解的作用:减少放射性气溶胶的质量;安全壳降温降压;安全壳注水。因此本文利用一体化严重事故分析程序,选取典型事故序列,评估防火喷淋系统在严重事故中的三种缓解作用的有效性为防火喷淋在严重事故管理导则中的应用提供技术支持。分析结果表明,防火喷淋系统能够实现堆腔淹没,在一定时间内进行安全壳降压,以及减少安全壳中放射性气溶胶的含量的作用,但由于系统限制,防火喷淋进行堆腔淹没的流量不能满足安全限值,并且只能推迟而不能够避免安全壳的失效。防火喷淋系统对严重事故的缓解作用虽然是有限的,但可为其他相关系统或设备的修复提供一定时间。 相似文献
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Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology 总被引:1,自引:0,他引:1
Muhammad Hashim Hidekazu Yoshikawa Takeshi Matsuoka 《Journal of Nuclear Science and Technology》2013,50(4):526-542
The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. 相似文献
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Tapani Kukkola 《Nuclear Engineering and Design》1988,109(1-2)
A modified Loviisa 1 & 2 plant, with a VVER-440 type of PWR, was designed to be an alternative for the fifth nuclear power plant in Finland. The designed plant is equipped with a full pressure double containment. The safety systems are in three separated safety trains. The plant will have one 500 MWe turbine. The newest design criteria were applied in the design. 相似文献
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The fire spray system (FSS) of the Advanced Passive PWR, as a part of the fire protection system, can provide a non-safety related containment spraying function for severe accident mitigation which is included in the Severe Accident Management Guidelines (SAMG) of the Advanced Passive PWR when dealing with severe accidents. The effectiveness of the FSS is investigated on three effects for severe accident mitigation which are controlling the containment condition, washing out fission product and injecting into the containment through three representative severe accident scenarios analysis with integral accident analysis code since there is no sufficient data support, besides the negative impact is also discussed. Results show that the FSS can be effective for controlling the containment condition, washing out fission product and injecting into the containment, however the effect is limited due to system limitation: the FSS can only cool the containment atmosphere for a short term; the flow rate of FSS cannot fulfill the success criteria given in the PRA report of the Advanced Passive PWR. Meanwhile, the hydrogen concentration and the containment water level should be the long-term monitored because actuating the FSS may cause hydrogen risk in the containment and containment flooding. Despite its limitation and negative impact, the FSS can be effective as an alternative severe accident mitigation measurement for postponing the process of accidents for safety system recovery. 相似文献
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快堆非能动安全研究发展概况 总被引:1,自引:1,他引:0
近年来,在美国快堆研究计划中,提出了一种实现快堆安全目标的新概念,就是以“纵深防御”思想为基础的“非能动安全”(Passive Safety)概念,强调应用非能动的机理保护反应堆和公众的安全,而不是依靠增加能动的专设安全设施。本文扼要介绍有关快中子反应堆“非能动安全”研究的发展概况。 相似文献
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Mirela Gavrilas Neil E. Todreas Michael J. Driscoll 《Progress in Nuclear Energy》1998,32(3-4):647-655
A containment is proposed for a high rating PWR (1300 MWe) that makes it possible to reject sufficient heat to maintain internal conditions below design limits during any postulated design basis accident. The proposed containment thus eliminates the need to employ active features for containment cooling, and conformes to guidelines set forth for passive reactor systems. [EPRI, 1987] The design is based in part on a currently operating PWR containment (Waterford 3). A series of modifications and additions are necessary to make passive heat rejection possible. The modifications are an increase in free volume and primary shell surface area. The additions are the perforation of the secondary containment structure to form an air-convection annulus, and allow the submersion of the lower part of the containment into an external pool; an internal pool increases in-containment heat storage. Proposed features are evaluated analytically, computationally and, where possible and necessary, experimentally. The proposed containment is shown to remain below current regulatory limits for the design-basis postulated loss of coolant accident. 相似文献