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姚进国 《核工程研究与设计》2005,(2):41-45
WWER-1000反应堆无论是在俄罗斯还是在国际上都算是比较成熟的堆型,它和西方PWR反应堆都属于压水堆的范畴。但在堆芯设计的理念上,WWER-1000与PWR堆之间又存在着差异。本文结合WWER-1000反应堆的运行实践及其燃料组件的特点,来阐明燃料棒弯曲在WWER-1000反应堆堆芯设计,尤其是在堆芯热工设计中是如何考虑的,并使之与西方PWR的设计特点进行针对性比较,从而来说明WWER-1000燃料组件的性能在整个运行寿期内的适应性。 相似文献
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压水堆核电站堆芯集中参数模型的微机仿真 总被引:1,自引:1,他引:0
阐述了PWR核电站堆芯的模型化问题,提适用于微机仿真的核电站堆芯的物理数学模型,将核电站堆芯分为三大块分别建立模型,中子动力学模块,反应性反馈模块,堆芯热力学模块,建立系统传递函数,运用MATLA仿真,得到良好结果。 相似文献
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中广核CPR1000核岛堆芯概念设计和安全裕度评估初探 总被引:1,自引:0,他引:1
CPR1000压水堆核电站是中广核集团20多年来经过渐进式改进和自主创新形成的中国改进压水堆核电站。CPR1000的参考设计是岭澳II期核电站加改进设计。在未来的10~15年内,CPR1000将是中广核集团主要建设的核电站类型之一。CPR1000的初始堆芯设计采用什么样的装料方式和燃料循环方式是必须首先解决和确定的重要设计前提,这是整个核岛设计、安全分析核执照申请的核心和基础。基于大亚湾核电站和岭澳核电站多年的燃料管理经验和运行经验以及国外类似核电站运行和设计经验,并且综合考虑了初始堆芯的特点和难点,以及不同堆芯设计和燃料管理策略的特点,对CPR1000的初始堆芯进行了设计。通过初步研究,本文提出了CPR1000初始堆芯采用的燃料组件类型,分析CPR1000采用从首循环开始进行18个月换料过渡的堆芯设计技术方案,并对CPR1000首循环实施18换料进行了堆芯设计安全裕度初步分析与评估。 相似文献
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压水堆核动力厂控制系统可视化模化与仿真软件——NCS 总被引:1,自引:0,他引:1
《中国核科技报告》1999,(1)
对压水堆核动力厂控制系统的建模与仿真进行了较系统的研究。提出了用于压水堆核动力厂控制系统快速和精确仿真的被控对象数学模型和数值方法。被控对象模型主要包括堆芯、稳压器、蒸汽发生器、管道及泵的模型等,分别采用了龙格-库塔法和特雷纳方法求解这些模型。设计了面向控制系统结构图的可视化模化平台,实现了图形方式下的控制系统可视化建模,并采用离散相似法对所建立的控制系统数学模型进行求解。研制出了相应的压水堆核动力厂控制系统可视化模化与系统仿真软件——NCS,并用NCS软件对商用核电站控制系统进行了仿真研究,得到了满意的结果。NCS软件对核电站控制系统设计和分析研究工作具有很好的参考和实用价值。 相似文献
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文中主要介绍了中国AC-600核电站在设计参数、堆芯设计、主系统、非能动专设安全设施等方面的设计特点,并与美国AP-600核电站设计进行了比较,给出了AC-600与AP-600的主要区别。 相似文献
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文中主要介绍了中国AC600核电站在设计参数、堆芯设计、主系统、非能动专设安全设施等方面的设计特点,并与美国AP-600核电站设计进行了比较,给出了AC-600与AP-600的主要区别。 相似文献
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The design philosophy and requirements of the HTR-10 reactor building and the primary loop confinement are introduced in this paper. Also introduced are the design, fabrication and the installation of the HTR-10 primary loop pressure boundary system. The primary loop confinement comprises the sealed cavities of the reinforced concrete structure. The main components and the connected gas systems of the primary loop pressure boundary system are contained in the confinement. Under normal operating condition, the inside pressure of the confinement is kept at negative pressure to ensure the sealing function of the confinement. There is a rupture disk of overpressure protection in the confinement wall. After a depressurization accident the pressure of the confinement increases and the rupture disk will break. The air of the confinement is discharged directly to the atmosphere through the accident discharge chimney which is connected to the rupture disk without filter. The main components of the primary loop pressure boundary system consist of the reactor pressure vessel, the steam generator pressure vessel and the hot gas duct vessel. All the above main components are installed in the reactor cavity and the steam generator cavity. They are all nuclear safety class 1 components, whose materials production, design, fabrication, and tests are carried out according to ASME Section III and relevant Chinese nuclear codes. 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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由于核电厂安全水平要求的逐渐提高,越来越多的非能动系统被用于先进反应堆堆型中,但对这些非能动系统可靠性评价的工作还处于初级阶段。本文根据非能动系统可靠性评价流程,通过RELAP5热工水力学程序模拟非能动系统物理过程,对AP1000反应堆压力容器外部冷却(ERVC)系统进行了可靠性评价。通过计算得到了压力容器下封头温度等参数的累积密度分布曲线,根据不同的成功准则即可获得AP1000 ERVC系统的可靠性。该非能动系统可靠性评价结果可用于核电厂PSA模型中,以更好地指导核电厂设计及提高核电厂的安全性。 相似文献
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The steam generators for the Fort St. Vrain nuclear power plant are the first application in the United States of once-through boiler design coupled with a high temperature gas-cooled reactor. They contain many design features which are unique for this type of component. Since they are an integral part of the primary system and completely enclosed by the prestressed concrete reactor vessel, they must be removable as well as fit the space available for penetrations. These requirements made the once-through boiler principle a logical choice. Multi-start helically wound tubes supported by perforated plates in a star-shape arrangement resulted in an extremely compact design. The helium inlet temperature of 1427°F and steam temperatures of 1005°F main and 1001°F reheat required unique solutions in terms of flexibility and cooling of support systems and selection of insulation materials and design. Operation in a helium atmosphere without a protective oxide layer called for materials with good wear protection characteristics where parts may experience relative motion. Stabilizing orifices, externally adjustable at the steam generator inlet, plus essentially equal tube lengths for each of the many parallel circuits are utilized to balance circuit performance. To minimize gas bypass flows, special gas seals are provided around individual tube bundles. Field erection time was minimized by developing an upper and a lower module assembly and joining them after erection in the reactor vessel. 相似文献
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在AP1000核电厂的某些严重事故情景中,安全壳可能发生失效或旁通,导致大量放射性物质释放到环境中,造成严重的放射性污染。针对大量放射性释放频率贡献最大的3种释放类别(安全壳旁通、安全壳早期失效和安全壳隔离失效),分别选取典型的严重事故序列(蒸汽发生器传热管破裂、自动卸压系统阀门误开启和压力容器破裂),使用MAAP程序计算分析了释放到环境中的裂变产物源项。该分析结果为量化AP1000核电厂的放射性释放后果和厂外剂量分析提供了必要的输入。 相似文献
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F. M. Mitenkov B. A. Averbakh B. A. Vasil’ev B. M. Kamashev K. L. Suknev 《Atomic Energy》2005,98(6):375-383
The results of design analyses for improving nuclear plants with fast reactors, specifically, by using cartridge-vessel generators instead of sectional-modular generators, are presented. It is shown for a nuclear power plant with a BN-800 reactor that the cartridge-vessel steam generators designed by the Special Machine Design Office substantially decrease the metal content, dimensions, mass, amount of construction work, and construction costs of the main vessel of the nuclear power plant.In the BN-800 design, a cartridge-vessel steam generator decreases the specific capital costs for constructing a power-generating unit of a nuclear power plant by approximately 8%, which substantially closes the gap between these costs for nuclear power plnats with BN-800 and VVER-1000 reactors.__________Translated from Atomnaya Energiya, Vol. 98, No. 6, pp. 403–412, June 2005. 相似文献
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An experimental study is performed to investigate the effects of noncondensable (NC) gas in the steam condensing system. A vertical condenser tube is submerged in a water pool where the heat from the condenser tube is removed by boiling heat transfer. The design of the test section is based on the passive condenser system in an advanced boiling water nuclear power reactor. Data are obtained for various process parameters, such as inlet steam flow rate, noncondensable gas concentration, and system pressure. Degradation of the condensing performance with increasing noncondensable gas is investigated. The condensation heat transfer coefficient and heat transfer rate decrease with noncondensable gas. The condensation heat transfer rate is enhanced by increasing the inlet steam flow rate and the pressure. The condensation heat transfer coefficient increases with the inlet steam flow rate, however, decreases with the system pressure. For the condenser submerged in a water pool with saturated condition, the strong primary pressure dependency is observed. 相似文献