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CEFR主容器内正弦三波激励下液面晃动响应 总被引:2,自引:0,他引:2
开发了一套可用于估算正弦三波激励下液面晃动对容器壁和顶盖冲击压力的工程方法,计算结果为中国实验快堆(CEFR)主容器及堆内构件的应力分析提供了重要的载荷输入。 相似文献
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中国实验快堆(CEFR)是钠冷快中子反应堆,其一、二回路的运行特性对反应堆的安全运行具有重要的影响。使用JTopmeret软件建立CEFR一、二回路主冷却系统和蒸汽发生器(SG)的仿真模型,用于计算系统任意一点的流量、压力、温度等运行参数。在稳态及瞬态工况下,系统主要参数仿真值与设计值的误差均小于2%,满足系统仿真的精度要求。 相似文献
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自然循环能力是衡量钠冷快堆固有安全性的重要指标,堆芯布置、回路设计及工况参数等都会影响堆芯自然循环能力,因此不同堆型的自然循环能力有很大差异。为了保证堆芯事故得到有效缓解,中国实验快堆(CEFR)的设计中通过优化系统布置,重点考虑了堆芯自然循环。本文采用SAS4A程序对CEFR进行系统建模,分析了CEFR在无保护失流(ULOF)工况下的堆芯热工水力参数瞬态特性,验证了CEFR利用自身自然循环和负反馈设计进行事故缓解的能力,本文还对一回路流动阻力和二回路钠装量对堆芯自然循环的影响进行分析。计算结果表明,CEFR具有良好的自然循环特性,在ULOF工况下可以依靠其负反馈停堆,并能够建立起稳定的自然循环从而导出堆芯余热。 相似文献
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堆顶固定屏蔽在中国实验快堆(CEFR)中承担着重要功能,对其进行充分冷却极其重要。本文采用CFD方法对该设备的冷却系统进行了三维数值研究,详细分析了该冷却系统的流动特性和水力学设计,并对设计中的不足提出了优化建议。研究表明,该冷却系统基本可满足要求,但部分环节需要优化。将调节阀尽量均匀布置可改善水平风道流场分布;入口处设置两道通风孔可提高竖直风道内空气流动的均匀性;调节阀开度应适当增加以进一步满足流量分配需求。该研究可为CEFR运行安全和类似冷却系统的设计提供参考。 相似文献
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CEFR的蒸汽发生器检测管由俄罗斯进口的10X2M钢与国产304L不锈钢焊接而成,其焊接区处在310℃流动钠中。诸多的快堆运行经验表明,部件的焊接处往往是部件受损的薄弱环节,因此,焊接件在高温钠中的抗晶间腐蚀性能及其相容性特征对CEFR的安全运行与安全分析有十分重要的意义。为确保快堆安全,本工作在CEFR的工况条件下进行了模拟试验,观察并分析焊接件材料与310℃钠的相容性特征及晶间腐蚀倾向,以为CEFR的安全运行及安全分析提供试验根据。 相似文献
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The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation. 相似文献
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The goal of the safety design for the demonstration fast breeder reactor is to ensure that the safety level is equivalent to or higher than that of the light water reactors of the same period. The design of the safety features such as reactor shutdown, decay heat removal and confinement systems is of importance to reach the goal. The reactor core is equipped with two independent fast shutdown systems, the primary system and the backup system. In addition, it is planned to strengthen the passive shutdown capability by using self- actuated systems such as a Curie point device for the backup system. The decay heat is removed from the core to the atmosphere through the safety lines of the direct reactor auxiliary cooling system which is composed of four independent lines. Furthermore, under the severe conditions that no active function of the decay heat removal system is available, the heat can be removed by natural convection through the safety lines by taking advantage of the high boiling temperature of sodium. For the confinement function, the reactor vessel is surrounded by a containment vessel and a confinement area.
The design concept of these safety features is described in this paper. 相似文献
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Neil E. Todreas Pavel Hejzlar Robert Petroski C.J. Fong M.A. Elliott 《Nuclear Engineering and Design》2009,239(12):2582-2595
Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO2) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400 MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. The S-CO2 reactor is directly coupled to the S-CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO2 uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material - materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U-TRU-Zr (10% Zr) for unity conversion ratio and TRU-Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr-1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development. 相似文献
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For a modular reactor of 200 MW thermal output an inactive after heat removal system has been designed. It consists of a prestressed cast iron pressure vessel with the surrounding reactor cell. Integrated in the cast iron profiles of the reactor cell is a redundant water cooling system based on natural convection. Air cooling towers are provided to cool the water down to ambient temperature. The cooling system covers a wide range of possible wall temperatures without significant changes in water temperature. The structures of the reactor pressure vessel and the cell, their assembly and some results of the engineering work done up to now are described in this paper. 相似文献
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Ayako Ono Hideki Kamide Jun Kobayashi Norihiro Doda Osamu Watanabe 《Journal of Nuclear Science and Technology》2016,53(9):1385-1396
A fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad range of plant operation conditions. The sodium experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, they were conducted by varying the pressure loss coefficients of the loop as the experimental parameters. These experiments confirmed robustness of the PRACS, which the increasing of pressure loss coefficient did not affect the heat removal capacity very much due to the feedback effect of natural circulation. 相似文献
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我国非能动系列压水堆将应急冷却系统冷却水的注入管道直接连接于压力容器上,与传统的冷管段安注不同,这种安注方式被称之为反应堆压力容器直接安注。本文以安注条件下的反应堆压力容器为研究对象,采用物理实验与数值分析结合的方法,对安注流体在压力容器表面形成的热分布形态进行研究。研究发现,不同于传统的主管道冷段斜接管安注方式,直接安注条件下安注流体在下降环腔中的分布形态接近于等腰三角形。以实验结果为基础,结合数值计算验证,发现了压力容器热分布角与流速比成正比关系,并进一步提出了安注流体分布计算模型,从而为反应堆安全设计提供参考。 相似文献