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核电厂蒸汽发生器主给水管道横跨设备冷却水系统(CCS)泵厂房,其中布置有柴油机、泵等重要设备。在CCS泵厂房发生蒸汽发生器主给水管道双端破裂事故工况下,需保证布置在CCS泵厂房内的CCS泵组不会因为水淹而造成失效,因此,需要对该漫流特性进行评价分析。已有研究大多关注管道破裂后流体高速喷射行为,而较少研究喷射流体在CCS泵厂房中漫流积淀情况,同时由于设备冷却水系统泵厂房空间尺寸巨大、结构复杂,很难开展原型尺寸实验研究。因此分别对破管位置位于CCS泵厂房5.334 m层空间和CCS泵厂房11墙与近核岛侧防甩墙之间的压力隔间两类事故场景分别进行三维数值计算。模拟结果表明:在蒸汽发生器双端断裂触发跳泵事故下,泄放水流量在11 s内即迅速下降,破口位置处于5.334 m层空间和压力隔间两类条件下均不会淹没CCS泵防水台,不影响CCS泵的正常运行。破口位于5.334 m层空间位置时设计预留开孔能有效排出漫流的泄放水;破口位于压力隔间内时设计的钢格栅也能有效排出漫流的泄放水。本研究为CCS泵厂房空间设备冷却水系统泵厂房防水淹策略优化设计提供重要数值参考。 相似文献
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在对放射性废液蒸发处理系统进行调试过程中,通过调节废液上料量、蒸汽发生器液位、一次蒸汽流量等系统参数来改变系统运行工况,得出各工况下的净化系数,分析系统净化效果的影响因素。调试结果表明:对于该系统,蒸汽发生器液位在500mm时净化系数最高;蒸发量为1m3/h时,净化系数最高;系统在变工况运行时产生波动,净化系数降低。系统原有两条控制联锁,为一次蒸汽流量与预热器出口温度、一次蒸汽流量与蒸汽发生器液位的联锁,仅此两条联锁对于系统的稳定性不够,且一次蒸汽控制液位的控制方式灵敏性差,滞后严重。文章通过分析系统运行各参数的关系,从系统运行稳定性和净化效果的角度,提出对该系统控制方式的合理改进——调整蒸汽发生器液位与上料量联锁控制。 相似文献
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CAP1000核电厂全功率范围SGTR事故研究 总被引:2,自引:2,他引:0
对CAP1000非能动核电厂在部分功率、零功率和热备用条件下发生的蒸汽发生器传热管破裂(SGTR)事故进行蒸汽发生器满溢评价。对典型的部分功率、零功率和热备用运行条件下的SGTR事故分别进行横向敏感性分析,选取每个运行条件下的保守工况,结合满功率事故工况进行纵向功率谱对比,根据其瞬态特性,分析事故进程,评价极限运行工况和关键参数。结果表明:CAP1000核电厂在全功率范围内发生SGTR事故均不会导致蒸汽发生器满溢,且最严重的工况发生在满功率条件下。 相似文献
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《Fusion Engineering and Design》2014,89(9-10):1975-1978
The ITER Tokamak requires multiple auxiliary systems to initiate, support, and monitor the fusion reaction. Heat produced by these systems, as well as the heat produced by the fusion reaction itself is collected by the ITER Cooling Water System (CWS) and rejected to the atmosphere. The CWS is composed of several systems designed for specific cooling roles. One of these systems is the Component Cooling Water System 2 (CCWS-2) whose function is to collect the heat from auxiliary client systems and components and transfer it to the Heat Rejection System. Clients are located throughout the site and have different requirements in terms of pressure, temperature, temperature variation, flow, metallurgy of wetted surfaces, and water quality. To satisfy these different requirements the CCWS-2 is divided into four separate loops, each of which has different operating parameters. For example, the CCWS-2A loop is designed to cool components with wetted surfaces of copper and primarily serves the radio-frequency heating systems, magnet power supplies, and neutral beam injector system components. This paper describes the evolution of the CCWS-2 system to match the needs of groups of compatible clients, and describes the development of the preliminary design of one of its loops, CCWS-2A, to meet individual client needs. 相似文献
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Cooling efficiency during transient reflooding under loss of normal coolant conditions has been examined with a 7 × 7 simulated fuel rod bundle and jet pump bypass. The bundle contains 49 electrically heated rods with 3600 mm heated length and a pseudo cosine axial power distribution. Water is injected into the lower plenum and the superheated bundle is reflooded from the bottom with some flow diverted to the simulated jet pump bypass. The results show that effective cooling can be maintained. 相似文献
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为了研究氦氢冷却气体对黑腔系统温度场的影响,采用CFD数值模拟方法,计算了氘氚靶丸外表面最大温差与填充区域的气体流场随气压、氦气含量变化的规律。通过对冷却壁面施加壁温扰动函数,监测了靶丸外表面平均温度、最大温差随时间的波动。研究结果表明:提高氦氢混合气体的填充压力或减小氦气含量,使得黑腔上下部分冷却气体自然对流强度差异增大,导致靶丸外表面温度场均匀性恶化;但降低冷却气体中氦气含量使气体导热系数减小,比热容增大,使得冷却壁温扰动对靶丸外表面温度场均匀性的影响减弱。 相似文献
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AbstractThe purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions. 相似文献
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Cooling process which uses water for heat transfer is an essential factor in coal-fired and nuclear plants. Loss of cooling upset can force the plants to shut down. In particular, this paper reports visual inspections and metallurgical examinations on the failed SA210-A1 right-hand side (RHS) water wall tube of a coal-fired plant. The water wall tube showed the abnormal outer surface colour and has failed with wide-open ductile rupture and thin edges indicating typical signs of short-term overheating. Metallurgical examinations confirmed the failed tube experiencing higher temperature operation. Water flow starvation due to restriction inside the upstream tube is identified as the main root cause of failure. The findings are important to take failure mitigation actions in the future operation. Discussion on the typical problems related to the cooling process in nuclear power plants is also presented. 相似文献
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针对超临界水堆(SCWR)控制棒落入堆芯事件特点,采用堆芯三维瞬态性能分析方法,利用开发的SCWR堆芯三维瞬态物理-热工水力耦合程序STTA,建立SCWR堆芯落棒瞬态三维计算模型和分析流程,研究分析超临界水堆CSR1000在控制棒落入堆芯瞬态过程中的堆芯性能,分析评价落棒瞬态下CSR1000堆芯的安全性能。堆芯三维落棒瞬态分析表明,当落入堆芯棒束价值较高时,落棒初期堆芯功率下降较快,之后由于水密度的反应性反馈,堆芯功率缓慢回升至新的平衡,堆芯功率下降速率超过了停堆信号整定值,将触发保护停堆;当落入堆芯棒束价值较低时,由于水密度的反应性反馈,堆芯功率下降缓慢,堆芯功率下降速率未能达到停堆信号整定值,不能触发保护停堆。控制棒落入堆芯对堆芯轴向功率分布影响很小,高价值落棒导致的落棒区域燃料组件功率坍塌相对低价值落棒更明显。无论是高价值落棒还是低价值落棒,瞬态过程中最大包壳壁面温度均低于瞬态安全限值850℃。水密度的显著反应性反馈及必要的保护停堆措施能保证CSR1000堆芯在控制棒落入堆芯过程中的安全性能。 相似文献
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为获得高温钠热管传热性能,开展真空条件下钠热管启动性能和等温性能试验,获得了钠热管真空条件下启动速度与等温性能数据;开展强制冷却工况条件下传热性能试验,获得了钠热管声速限特性与试验工况下的最大传热功率。经试验验证,所研制高温钠热管在真空条件下,580 ℃时完全启动,启动用时20 min,轴向壁面温差低于11 ℃,等温性能良好;钠热管传热功率在工作温度为500~650 ℃时受声速极限限制,在650 ℃以上受携带极限限制;在750 ℃和850 ℃时,测得热管最大散热功率分别为4.78 kW与8.02 kW,对应的最大轴向热流密度分别为1.51 kW/cm2与2.53 kW/cm2。试验结果表明,所研制钠热管具有较强传热能力,可满足热管式核反应堆等工程应用需求。 相似文献
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本文基于我国聚变工程实验堆水冷包层优化设计与安全分析的要求,针对水冷包层模块第一壁的流动传热特性进行三维数值模拟研究。采用计算流体力学方法,建立了水冷包层模块第一壁的三维数值模型,研究流量分配的特点以及温度分布情况,分析与评估在稳态工况、瞬态工况及失流事故下的水冷包层模块第一壁传热能力。研究结果表明,不同冷却管间存在流量分配不均匀的现象;在稳态工况下,水冷包层模块第一壁具有较好的传热能力,瞬态工况下水冷包层模块能够有效地导出反应堆热量;失流事故下冷却管内温度短时间上升至系统压力下的饱和温度,有待进一步研究。相关研究为优化包层第一壁传热设计提供参考,并为今后聚变堆的安全分析提供依据。 相似文献
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中欧核能合作研究项目超临界水堆燃料验证实验(SCWR-FQT)的主要研究内容为在超临界水环境下对一个小型燃料组件进行堆内性能分析和验证。本文应用修过后的系统程序ATHLET-SC对该实验回路进行建模,同时结合堆芯中子物理的计算结果,对由于压力管进口管破裂形成的失水事故进行热工水力和中子物理的耦合分析,并讨论了物理耦合中停堆棒的负反应性、冷却剂温度系数等参数对结果的影响。计算结果表明,进行了中子物理耦合的结果得到的最高包壳温度比未进行中子耦合的结果要低15℃,同时停堆棒引入的负反应性是该事故过程中影响燃料棒最高包壳温度的一个主要因素。 相似文献