大破口LOCA喷放模拟实验启动方法研究

地坑滤网性能评价及下游效应分析中通常需要开展大破口失水事故(LOCA)喷放模拟实验。本研究采用双膜爆破片结合气动阀实现大破口LOCA喷放模拟实验启动压力的精确控制和高能流体的瞬间释放。压力测量结果表明:双膜放气启动压力损失最大,单膜启动压力爬升较慢,双膜充气启动在压力损失、压力爬升速度和压力精度上都能准确模拟大破口LOCA喷放物理过程。最后采用双膜充气启动方法对双壁盒式保温结构进行射流破坏试验,结果表明本装置提供的冲击力度足够。     

基于RELAP5的LOCA喷放阶段下降段内CCFL特性研究

反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL）,不利于应急冷却水顺利进入堆芯,极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为;通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性,并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明,由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性,随着蒸汽流量增大,在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大,较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系;在靠近破口的环路内注入的冷却水更难到达下腔室,而在远离破口环路的冷却水容易进入到下腔室;过热的蒸汽在流动过程中被冷却水冷却发生凝结现象,导致出口蒸汽流量小于进口蒸汽流量,且随着进口蒸汽流量的增大,凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。      

再淹没传热模拟的移动子网格模型

通过计算得知，在较严重的冷却剂丧失（LOCA）事故中，坍塌阶段后期，堆芯将会丧失有效冷却。反应堆堆芯在这个阶段经历了一个几乎是绝热的升温过程。冷水进入堆芯后与较高温度的燃料棒接触，开始重新建立有效的堆芯冷却。在这个阶段包壳温度达到最大值？当燃料棒淬火时将产生较高流率的过热蒸汽。蒸汽流速通常足够高，能够带走以液滴形式存在的较大份额的液体。     

喷淋液滴在中低压饱和蒸汽环境下传热特性研究

本文以喷淋液滴在中低压饱和蒸汽环境下传热特性为工程背景,分析液滴初始动力学参数、环境蒸汽参数对液滴与蒸汽间传热特性的影响。计算结果表明,不同液滴初始速度下单位质量换热量随时间变化趋势基本一致,初始速度越大,换热系数越高,单位质量换热量也越大,液滴表面温度和平均温度达到饱和蒸汽环境温度的时间越短;液滴直径对单位质量换热量随时间变化趋势影响较为显著,液滴直径越大,单位质量换热量随时间降低的趋势逐渐变缓,液滴表面温度和平均温度达到饱和蒸汽环境温度的时间越长;饱和蒸汽压力越大,液滴表面温度和平均温度上升越快,但不同蒸汽压力下液滴表面温度或平均温度达到饱和温度的时间基本一致;不同液滴初始温度时,液滴表面温度、平均温度、单位质量换热量随时间变化基本一致。计算结果有助于优化工程实际中喷淋系统的设计。     

基于ADS1～3缩比模型的双孔蒸汽喷放冷凝传热系统程序计算与实验验证

高温蒸汽在过冷水中喷放直接接触式冷凝是AP1000、CAP1400等三代先进压水堆一回路在事故超压情况下重要的降温降压途径。本文基于系统程序RELAP5、COSINE对饱和蒸汽通过双孔喷洒器喷入大容积过冷水中进行直接接触冷凝这一过程进行建模、计算、分析,获得高温蒸汽从喷口喷出后沿轴向的温度分布。同时开展蒸汽喷放冷凝可视化实验,采用热电偶矩阵和高速摄像机等对关键热工参数进行测量,以获得蒸汽汽羽的温度分布和喷放流型等,用于验证系统程序对蒸汽喷放冷凝过程模拟的准确性。结果表明,采用RELAP5程序基本能模拟简化条件下的ADS蒸汽喷放冷凝总体变化规律,模拟结果与实验结果相比平均误差为2.97%。此外,采用COSINE程序对喷放冷凝过程模型进行了进一步修正和改进,考虑水箱内整体流动对喷放特性的影响,模拟结果与实验结果吻合较好,平均误差为1.89%。但由于实际双孔喷放过程较为复杂,并且存在明显的三维特性,所以仍需对系统程序中相关冷凝传热模型进行完善,以更精确地模拟其局部冷凝特征。     

抑压式安全壳的抑压特性研究

以100 MW级核电厂压水堆为对象,通过对反应堆冷却剂失水事故(LOCA)初期安全壳压力温度响应的分析,对抑压式安全壳抑压特性进行研究。由于LOCA事故喷放阶段质能释放焓值较高,安全壳喷淋难以及时有效地抑制安全壳压力的上升,而采用抑压水池对抑制事故初期的压力具有较为明显的效果。通过对抑压水池总容积、气水容积比、排放管流通面积等重要参数的分析,对抑压效果的影响表现为:其中抑压水池总容积大小对抑压效果影响程度最大;并且抑压水池气水容积存在最佳比;排放管流通面积存在最佳范围。     

主泵参数变化对失水事故后果影响分析

基于100D主泵和ANDRITZ主泵的差异,分析主泵相似特性曲线和自由容积的变化对失水事故(LOCA)后果的影响。针对岭澳核电站二期反应堆冷却剂系统,应用CATHARE GB程序和CONPATE4程序分析大破口LOCA事故堆芯热工水力后果;应用ATHIS和FORCET程序分析失水事故喷放阶段的反应堆冷却剂主管道水力载荷。结果表明,主泵相似特性曲线的变化对大LOCA事故再淹没阶段的堆芯热工特性影响很大,采用不同主泵时的最高峰值包壳温度(PCT)相差很大;而主泵自由容积对失水事故喷放阶段的卸压波传递影响较大,导致采用不同主泵时的反应堆冷却剂主管道水力载荷有所不同。     

LOCA下喷放水流对周围仪表仪器的影响分析

在核电厂一回路发生冷却剂流失事故(LOCA)时,冷却剂从破口喷出,急速汽化,可对周嗣的仪表仪器造成强冲击力的破坏,产生极为严重的后果.本文以900MW压水堆为研究对象,使用数值模拟的方法,建立一维喷放模型,分析LOCA可能造成的破坏力.分析结果表明,以激波作为分界线,在激波形成前的仪表仪器将受到强冲击力,而在激波形成后...     

核反应堆冷却剂丧失事故下喷淋液滴特性研究

在核反应堆发生冷却剂丧失事故（LOCA）时,大量质能释放到安全壳内,通过喷淋能有效降低安全壳内的温度及压力。考虑喷淋液滴下落过程中体积、质量、温度及阻力系数的变化,应用对流传热及传质关联式,计算得到液滴与周围介质的传热、传质特性。计算结果与TOSQAN试验对比符合较好。对不同液滴尺寸、不同环境压力及蒸汽未达到饱和的情况进行了计算,分析了喷淋的影响,结果可为喷淋系统的设计与应用提供一定的理论依据。     

小型压水堆抑压水池抑压特性研究

与传统大型压水堆相比，小型压水堆安全壳自由容积小，发生冷却剂失水事故(LOCA)后安全壳压力迅速上升，需采取抑压水池、安全壳喷淋等措施保证安全壳的完整性。为探究抑压式安全壳中抑压水池的抑压特性，设计了小型压水堆抑压喷淋系统模拟装置，并使用RELAP5程序对模拟装置进行建模，模拟安全壳抑压排热过程，分析了破口质能释放对抑压水池模拟和设计的影响，对比了不同破口喷放速度和喷淋流量下抑压水池采用不同抑压管面积和气水比的抑压效果。结果表明：抑压水池的抑压管面积和气水比存在最佳值；破口面积影响喷放速度，在质能释放总量相同的前提下，破口喷放速度越快，抑压水池的最佳抑压管面积和最佳气水比越大；增设喷淋后，最佳气水比随喷淋流量增大而增大，最佳抑压管面积在小范围内变化。研究结果可为小型压水堆抑压式安全壳的设计和分析提供参考。     

Numerical Study on Influence of Blowdown Parameter on Coolant Blowdown Characteristic in LOCA

The coolant blowdown process is one of the important processes of the loss of coolant accident (LOCA). It is of great significance to study the thermal hydraulic characteristics of coolant blowdown process for understanding LOCA and predicting the migration process of radioactive source term after accident. The numerical simulation model of coolant blowdown was established by FLUENT software and verified. The model was used to study the effects of blowdown parameters such as diameter of nozzle, blowdown distance and blowdown pressure on flow field temperature, droplet velocity and vapor velocity. The results show that the increase of diameter of nozzle increases the blowdown parameters. As the blowdown distance increases, the flow field temperature and the droplet velocity increase first and then decrease, while the vapor velocity first rises and then stabilizes. The greater the blowdown pressure is, the farther the blowdown parameter is from the blowdown outlet. The maximum values of droplet velocity and vapor velocity increase gradually with the blowdown pressure, while the maximum value of the flow field temperature does not change.     

主泵两相降级对大破口失水事故的影响研究

大破口失水事故过程中,主泵的工作范围覆盖了单相液、气液两相和单相气工况。在两相工况下,主泵的扬程和转矩发生降级。对于AP1000核电厂,WCOBRA/TRAC被用于大破口失水事故分析,其现有的主泵两相降级数据来源于西屋W93A主泵。为正确模拟AP1000主泵在大破口失水事故过程中的热工水力特性,需对其两相降级特性进行研究。本研究分别采用国际上广泛使用的SEMISCALE和EPRI/CE主泵的两相降级数据进行AP1000冷段双端断裂事故的计算分析,并与原有W93A的计算结果进行对比。结果表明,AP1000主泵两相降级特性对反应堆冷却剂系统压力、破口流量和安注箱流量影响不大。相比于SEMISCALE和EPRI/CE,现有的W93A的两相降级数据将导致更低的堆芯冷却流量和更高的包壳峰值温度最大值,计算结果相对偏于保守。     

LOCA Analysis of Super Fast Reactor

This paper describes loss of coolant accident (LOCA) analyses of the Supercritical-pressure Water-Cooled Fast Reactor (Super Fast Reactor). The features of the Super Fast Reactor are high power density and downward flow cooled fuel channels for the improvement of the economic potential of the Super Fast Reactor with high outlet steam temperature. The LOCA induces large pressure and coolant density change in the core. This change influences the flow distribution among the downward flow parallel channels. It will affect the safety of the Super Fast Reactor. LOCA analysis of Super Fast Reactor is important to understand the safety features of the Super Fast Reactor. Keeping the flow rate in the core is important for the safety of the Super Fast Reactor. In LOCA, it is difficult to maintain an adequate flow rate due to the once-through coolant cycle and the downward flow cooled fuel assemblies. Therefore, the early actuation of the Automatic Depressurization System (ADS) and reduction of the maximum linear heat generation rates of the downward flow seed fuel assemblies and Low-Pressure Core Spray (LPCS) system are necessary for the Super Fast Reactor to cool the core under LOCA. Analysis results show that the Super Fast Reactor can satisfy the safety criteria with these systems.     

超临界CO2临界流稳态试验研究及模型验证

反应堆发生破口事故时,由于堆内处于高温高压状态而外界压力很低,破口处可能出现临界流动现象,临界流动特性对事故进程有较大影响,破口临界流量的准确估算对超临界水堆的安全分析更为重要。针对喷放为两相流动的工况范围,以超临界CO₂为工质,采用直径2 mm、长径比1～20的喷管试验段在超临界压力下开展了临界流稳态试验,获得了系统可靠的试验数据,研究了滞止压力、滞止温度以及喷管长径比对临界流量的影响。使用获得的超临界CO₂试验数据验证了临界流热平衡通用模型的通用性和准确性,发现其可以较好地预测超临界工况下的临界流量。本文研究补充了临界流动试验数据库,为临界流模型的验证和改进积累了试验数据。     

Blowdown thrust force under pipe rupture accident: II. Analytical evaluations of partial break and ramped opening break

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.     

Numerical Study of Depressurization Rate during Blowdown Based on Lumped Model Analysis

To study the thermodynamic aspects of blowdown, the depressurization rate equation has been numerically solved. The equation, derived from macroscopic mass and energy balances in the pressure vessel, consisted of the energy and volumetric discharges terms multiplied by the decrease rate of residual coolant. By applying a dimensional analysis, dimensionless equations were obtained together with dimensionless parameters of blowdown. Blowdown calculations starting at typical BWR operating conditions indicated that the decrease rate of coolant increased for the liquid and two-phase mixture, and decreased for the vapor discharge. Further, the energy discharge term made a larger contribution to the depressurization rate in the case of vapor escape, while the volumetric discharge term did so in the case of liquid and two-phase mixture escape blowdowns. In the lumped model analyses, the averaged specific enthalpy and entropy of the residual coolant increased for the liquid discharge, remained almost constant for the two-phase mixture discharge, and decreased for the vapor discharge blowdown.     

Blowdown thrust force under pipe rupture accident: I. Experimental evaluations of blowdown thrust force and decompression characteristics

Blowdown thrust forces and decompression characteristics were evaluated concerning the jet discharge or pipe whip tests with a 4-inch or 6-inch diameter pipe under PWR LOCA or BWR LOCA conditions related to pipe rupture accidents in nuclear power plants. This paper presents experimental evaluations of time-dependent and maximum blowdown thrust forces, and evaluations of decompression characteristics under instantaneous pipe rupture conditions.The following items are discussed: the peak value of the blowdown thrust force, the jet thrust coefficient for the maximum blowdown thrust force, the pressure recovery after break, and the relationship between the pressure undershoot of the sudden decompression and the decompression rate.     

A study of the sensitivity of loca heat transfer analysis for a water-cooled reactor system

Results are given of computer calculations, using the reactor thermal analysis code THETA1-B, to determine the significance and relative importance of various heat transfer regimes in predicting maximum fuel cladding temperature for the blowdown phase of a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor system. The factors considered include the choice of heat transfer correlation for a particular heat transfer regime, the method of delineating the boundaries between regimes, and core inlet coolant flow conditions.For a hot-leg rupture, the maximum surface temperature is sensitive to a number of factors, including choices of critical heat flux correlation, flow boiling transition heat transfer correlation, and in particular, stable film flow boiling correlation. However, for a LOCA resulting from a double-ended rupture of an inlet feeder, these factors have only marginal effects on maximum cladding temperature. In this case the importance of heat transfer to dry steam coolant at low net flow rate conditions is demonstrated, indicating a need for further information.     

西安脉冲堆大破口失水事故分析

分析了西安脉冲堆大破口失水事故的特点,建立了适用的数学模型,编制了计算程序。结果表明：在大破口失水事故下,部分燃料芯体最高温度将超过设计限值,但不会发生燃料元件熔毁事故。     

延展表面凝结液膜的动量积分方法

采用动量积分方法分析压水堆发生失水事故时在安全壳的内表面上的液膜凝结、再浸润和蒸发过程。由凝结液膜的质量和动量守恒方程导出了凝结液膜在延展表面的子午线方向平均速度的积分-微分方程。假设液膜以层流的方式流动，把导出的积分-微分方程变成容易进行数值积分的液膜速度的一阶常微分方程，由此求得液膜厚度分布。液膜能量守恒方程的解给出了安全壳内壁面的温度分布。