共查询到18条相似文献,搜索用时 46 毫秒
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介绍了秦山CANDU6机组因考虑到CANDU9设计的新技术,对主控制室所作的设计改进。主要是大屏幕显示器和配套的电站显示系统,优化的CRT报警系统和美学设计改进。 相似文献
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对秦山三期CANDU6机组停堆大修时堆芯燃料过热的敏感性进行了分析,介绍了秦山三期CANDU 6核电站1#机组101大修期间热阱的安全原则、管理细则.对停堆期间不同的堆芯余热情况下,失去冷却热阱后燃料达到过热前允许的恢复热阱时间进行了分析计算;并结合首次大修对大修期间低水位工况下停堆冷却泵连续运行、失去四级电源、临时盖板的使用等问题进行了分析,提出了具体建议. 相似文献
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介绍了秦山三期重水堆核电站为适应中国国情和秦山厂址和根据CANDU技术与国际核工业的发展,对原CANDU6的设计所作的15项重要变更。 相似文献
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介绍了秦山三期重水堆核电站为适应中国国情和秦山厂址和根据CANDU技术与国际核工业的发展,对原CANDU6的设计所作的15项重要变更。 相似文献
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对秦山三期CANDU堆应用稍浓铀的可行性用DRAGON/DONUON程序做了时均堆芯研究分析确定秦山三期采用稍浓铀的最优富集度为1.125wt%并对使用此富集度稍浓铀的秦山三期CANDU堆做了基于通道年龄模型的瞬时堆芯检验计算结果表明,在使用2.4棒束换料及简单的分2个燃耗区,外内区燃耗比为0.9时,能够满足秦山三期运行执照限制秦山三期CANDU堆使用此富集度燃料的经济效益的初步分析表明,它将使燃耗提高到185GWd/t(U),每年节省天然铀资源约53吨,减少乏燃料约116吨,节省燃料循环费用约6700万元计算表明,勿需对秦山三期堆芯结构和运行模式做重大改造即可完成天然铀向稍浓铀的过渡。 相似文献
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针对CANFLEX组件装载MOX燃料在CANDU重水堆中的应用进行了时均和瞬态验证计算。计算结果表明,最大通道功率和最大棒束功率均未超过限值。勿需对堆芯结构和运行模式做重大改变即可完成从天然铀堆芯向MOX堆芯的过渡。提出了应用MOX燃料的PWR/CANDU联合燃料循环策略。估算表明,秦山三期CANDU堆采用先进PWR/CANDU联合燃料循环,将使燃耗提高到13900MW·d/t(U);相对于PWR和CANDU堆各自独立的燃料循环,每年节省天然铀资源180t,减少乏燃料处置量约128t。 相似文献
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我国秦山三期核电厂采用的是加拿大CANDU-6反应堆机组。这是我国首次引进重水压力管式反应堆堆型,为了满足这一新堆型燃料管理计算的需要,开发了CANDU堆燃料管理的计算软件DRAGON/DONJON。并采用这套程序对秦山三期CANDU-6反应堆进行了一些初步的燃料管理计算。许算结果表明DRAGON/DONJON可满足秦山三期核电厂燃料管理计算需要。 相似文献
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对CANDU6反应堆厂房主设备两种不同的安装模式从技术、经济、进度等方面进行了分析比较,说明在秦山三期工程中采用LR1650履带吊车从厂房顶部吊装主设备,技术上操作简便,安全可靠,进度上可缩短工期,经济上也给电站带来可观的效益。 相似文献
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CANDU6堆调试期间的物理启动和物理调试试验主要在B阶段和C阶段进行,同时也包括A阶段的堆芯装料后的临界监督。基于韩国月城4号机组的物理调试,介绍了物理调试试验数据的预模拟分析、试验方法以及试验结果的分析和评估,并指出了试验中的重要注意事项。 相似文献
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详述了秦山CANDU堆燃料操作系统,包括新燃料贮存和运输系统、乏燃料贮存和运输系统、燃料更换系统以及远距离观测摄像机。 相似文献
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In CANDU reactors, the cool moderator surrounding the calandria tubes provides a potential heat sink following an accident initiator if the emergency coolant injection fails. However, in scenarios when a subsequent loss of all heat sinks occurs, the fuel channels fail and ultimately, the entire reactor core collapses and relocates into the bottom of calandria vessel (CV), which is externally cooled by shield-tank water. Previous studies using MAAP4-CANDU and ISAAC computer codes were found to investigate the long-term coolability of the CV in the late phase of core degradation in course of a severe accident. SCDAP/RELAP5 was applied in a previous work of the authors to the study of the in-vessel retention issue using the COUPLE models with user-defined slumping inside the 2D COUPLE mesh. This option allows for thermal and mechanical analyses of the reactor lower head avoiding the necessity to calculate the preceding course of core degradation during the accident. The former analyses used an equivalent spherically shaped CV while, for the present paper, calculations are performed with COUPLE routines modified to properly use the option for a horizontal pipe in plane geometry. The paper describes the modifications and the application of the resulted SCDAP/RELAPSIM/MOD3.4 code version to the study of the coolability of a CV starting with a dry debris bed. The vessel rupture time is compared to the ISAAC calculated value for a LOCA with loss of all heat sinks and no recovery actions. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty: boundary conditions of the vessel above debris, gap heat transfer coefficient and metallic fraction of zirconium inside the debris. 相似文献
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CANDU堆是世界上达到充分成熟且成功发展的少数几种堆型之一。这种堆的设计概念的基本出发 是使用天然铀燃料,这一选择决定了其它几个有利的选择,例如采用重水慢化剂,不停堆换料以及计划机控制。采用重水慢化剂和增大输出功率的需求决定了压力管式堆芯结构的设计方案。 相似文献
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An investigation on recycling the recovered uranium from electro-refining process in a CANDU reactor
Feasibility studies for recycling the recovered uranium from electro-refining process of pyroprocessing into a Canada Deuterium Uranium (CANDU) reactor have been carried out with a source term analysis code ORIGEN-S, a reactor lattice analysis code WIMS-AECL, and a Monte Carlo analysis code MCNPX. The uranium metal can be recovered in a solid cathode during an electro-refining process and has a form of a dendrite phase with about 99.99% expecting recovery purity. Considering some impurities of transuranic (TRU) elements and fission products in the recovered uranium, sensitivity calculations were also performed for the compositions of impurities. For a typical spent PWR fuel of 3.0 wt.% of uranium enrichment, 30 GWD/tU burnup and 10 years cooling, the recovered uranium exhibited an extended burnup up to 14 GWD/tU. And among the several safety parameters, the void reactivity at the equilibrium state was estimated 15 mk. Additionally, a simple sphere model was constructed to analyze surface dose rates with the Monte Carlo calculations. It was found that the recovered uranium from the spent PWR fuel by electro-refining process has a significant radioactivity depending on the impurities such as fission products. 相似文献