共查询到19条相似文献,搜索用时 375 毫秒
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提出了一种可应用于钍基先进CANDU型反应堆(TACR:Thorium-based Advanced CANDU Reactor)压力管与排管间的非能动热开关设计方案.该方案应用金属的热胀冷缩性质,通过热胀冷缩部件推动开关滑块移动来控制压力管与排管间的传热介质种类,以改变压力管与排管之间的热阻.该方案在满足TACR正常运行工况下对压力管和排管间高热阻要求的同时,能够在事故工况下降低二者之间的热阻导出余热.由于利用了金属热胀冷缩性质作为推动力,并利用改变传热介质种类来改变热阻,因此,高度的可靠性和有效性是该方案设计的特点. 相似文献
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在气冷CANDU式燃料组件之中,辐射换热也是不容忽视的一部分。特别是在出现了系统失压/失流事故时,辐射换热将会成为保证燃料安全的主要冷却手段。本文中针对CANDU式压力管编制了针对压力管几何条件下的一维辐射换热瞬态程序。程序中采用将燃料元件棒转化为同心圆环的方式简化辐射角的计算,并加入了隔层辐射模型,使模型更加贴近实际。采用分别将程序中的几个模块的计算结果与CFX计算结果对比的方式来达到程序验证的目的,验证结果显示程序RHTPB具有良好的表现,能够满足于反应堆安全计算的需要。 相似文献
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《核动力工程》2017,(6):51-56
采用带移动边界的三维瞬态模型对1/4堆芯模型进行热传导分析。考虑了堆芯熔融物滞留工况下反应堆衰变功率和压力容器内部水位的下降过程,以及不同材料组件在堆内的真实径向分布。棒束表面与冷却剂的自然对流换热采用饱和蒸汽/水经验关系式计算,辐射换热采用相邻16棒间辐射模型计算。建立了动态烧蚀模型以模拟不断累积的堆芯熔融物对下支撑板的烧蚀作用。着重考虑了由熔融物滴落造成的冲击换热以及下支撑板上形成的熔坑底部换热。文献验证对比证明了模型的正确性。模拟结果表明:事故进程2600 s时,冷却剂蒸干造成堆芯融化速度急剧加快。8000 s时80%的堆芯质量熔化。下支撑板上的烧蚀区域主要集中于板心半径700 mm处,并在6000 s时完全贯穿。 相似文献
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针对石墨慢化通道式熔盐堆的堆芯结构,基于COMSOL Multiphysics程序和MATLAB程序建立了堆芯稳态热工水力学计算模型。该模型对堆芯内固体区域的温度分布采用三维热传导方程进行模拟,对通道内熔盐温度采用一维单相流体模型进行计算。固体区域与熔盐通过熔盐通道壁面的对流换热边界建立热耦合。该模型基于平行通道压力损失相等的原则,分配堆芯内各熔盐通道的流量。通过对比RELAP5程序的计算结果,验证了模型对温度和流量分配计算的正确性。针对2 MWt 液态燃料熔盐堆的一种概念设计,分析了堆芯内三维温度分布和通道间流量分配。该模型可精确计算通道式熔盐堆堆芯内稳态温度分布和流量分配,对堆芯的热工水力学设计具有重要意义。 相似文献
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本文针对可应用于空间堆的矩形液滴辐射器(LDR),研究其液滴流的辐射换热及蒸发特性。在传统的高温液滴流辐射换热模型的基础上,添加了液滴流蒸发模型,并将辐射换热模型与蒸发模型进行耦合,在该模型的基础上开发了高温液滴流辐射换热-蒸发特性分析程序LDFAC。将该程序的液滴层温度分布计算结果进行校核,其相对误差不超过1.9%。使用该程序对装载DC705硅油下不同光学厚度及长度的液滴层辐射换热蒸发特性进行了分析。结果表明:在液滴层的光学厚度较大的情况下,液滴层内部的温度分布非常不均匀,液滴层中心的温度没有显著降低,而液滴层接近外表面部分的温度下降较为明显;温度对LDR的系统寿命有着较大影响,温度每降低10 K,系统寿命可提高约450%,同时,液滴层光学厚度越大,系统寿命也越长。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(7):726-734
In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure tube, and the same between pressure and calandria tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe2O3 powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure tube made of Zr-2.5%Nb and calandria tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental results without crud compare favorably with Tachibana's correlation. The experimental contact conductances with crud are found to be lowered. Since Tachibana's correlation was not applicable to the present situation, the empirical correlation was proposed. The correlation fitted well with the experimental values. 相似文献
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In the case of a loss-of-coolant accident (LOCA) with coincident loss of emergency coolant injection (LOECI), core cooling is generally very severe. However, as the ATR plant has heavy water at about 60°C in the core, decay heat can be removed by the heavy water cooling system. Separate-effects tests relating to heavy water cooling were conducted with each setup. The important thermal hydraulics was radiation heat transfer, ballooning of a pressure tube, contact conductance between the pressure tube and a calandria tube and critical heat flux of the calandria tube. Constants and correlations obtained by the tests were incorporated into several codes to assess the core cooling. Long term core cooling capability with the heavy water cooling system was assessed. The core was cooled without melting under the postulated events due to inherent characteristics of the ATR. 相似文献
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Gheorghe Negut Alexandru Catana Ilie Prisecaru Daniel Dupleac 《Annals of Nuclear Energy》2009,36(9):1424-1430
Romania as UE member got new challenges for its nuclear industry. Romania operates since 1996 a CANDU nuclear power reactor and since 2007 the second CANDU unit. In EU are operated mainly PWR reactors, so, ours have to meet UE standards. Safety analysis guidelines require to model nuclear reactors severe accidents.Starting from previous studies, a CANDU degraded core thermal hydraulic model was developed. The initiating event is a LOCA, with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperature inside a pressure tube reaches 1000 °C, a contact between pressure tube and calandria tube occurs and the decay heat is transferred to the moderator. Due to the lack of cooling, the moderator, eventually, begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) uncover, then disintegrate and fall down to the calandria vessel bottom. All the quantity of calandria moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which surrounds the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. 相似文献
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采用一体化分析程序建立了包括热传输系统、慢化剂系统、端屏蔽系统、蒸汽发生器二次侧系统的重水堆核电厂的严重事故分析模型。并选取出口集管发生双端剪切断裂的大破口失水事故(LLOCA),同时叠加低压安注失效,辅助给水强制关闭的严重事故序列进行热工水力分析。由于主热传输系统环路隔离阀的关闭,使得两个环路的热工水力响应过程不同。最终由于低压安注的失效,慢化剂系统逐渐被加热,最终导致堆芯熔化、排管容器蠕变失效。在LLOCA事故序列中叠加向排管容器中注水的缓解措施,可以终止事故进程,使堆芯保持安全、稳定的状态。 相似文献
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For the blind calculation of the International Collaborative Standard Problem (ICSP) experiment on heavy water reactor moderator subcooling requirements, the COMSOL Multiphysics code is used to simulate plastic deformation of a pressure tube (PT) as a result of the interaction of stress and temperature. It is shown that the thermal stress model of COMSOL is compatible to simulate the multiple heat transfers (including the radiation heat transfer and heat conduction) and stress strain in the simplified two-dimensional problem. The benchmark test result for radiation heat transfer is in good agreement with the analytical solution for the concentric configuration of PT and calandria tube (CT). Since the original strain model of COMSOL only considers an elastic deformation with thermal expansion coefficient, the PT/CT contact cannot be predicted in the ICSP. Therefore, the plastic deformation model by the Shewfelt and Godin, widely used in the fuel channel analysis of CANadian Deuterium Uranium (CANDU) reactor, is implemented to the strain equation of COMSOL. The heat-up of PT, the strain rate, and the contact time of the PT/CT are calculated with the boundary conditions (BCs) given for blind calculation of the ICSP experiment.
The result shows a sudden expansion of the inner concentric PT within a few milliseconds. This unsteady simulation should be helpful for the conceptual design of experiment as well as for the understanding of multiphysics inside the fuel channels of the CANDU reactor. 相似文献
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泵致脉动压力是核电站中引起主设备部件疲劳失效的主要原因之一。本文建立了蒸汽发生器传热管的泵致脉动压力载荷表达式,并建立不同弯曲半径的传热管有限元模型,对蒸汽发生器传热管在泵致脉动压力载荷下的动力学响应进行了研究。结果表明:34、64、94、114、124、144排传热管附近的频率、振型对泵致脉动压力最为敏感;包络泵致脉动压力作用下,最大应力出现在32排传热管上;传热管在泵致脉动压力载荷作用下,泵致脉动压力载荷的轴频频率对结构响应的贡献最大。本文分析结果为蒸汽发生器传热管在泵致脉动压力载荷下的磨损分析提供了参考。 相似文献