小型核动力装置抽真空启动过程热力特性及含氧量变化实验研究

在压水堆核动力装置启动阶段采用抽真空的方式对一回路系统进行抽气除氧,可以控制一回路冷却剂含氧量、减缓材料腐蚀并加快启动速度。为研究小型核动力装置一回路系统的抽真空启动特性,设计并搭建了小型核动力装置抽真空启动实验系统,通过实验获得了回路在抽气、注水、建立汽腔及升温升压过程中的温度、压力以及含氧量变化规律。结果表明：对于小型核动力装置,采用抽真空方法可以实现半小时左右完成抽气及注水过程,且回路冷却剂的溶解氧低于0.1×10^-6(质量分数);在抽真空过程中达到的真空度越高,启动过程中回路内冷却剂含氧量越低;通过分析启动过程中不同抽真空压力下的回路冷却剂含氧量,发现在水装量较小的核动力装置中,回路中未溶解的氧占有较大比重,需要进一步对氧气溶解的瞬态过程进行分析。     

一回路剩余空气体积对核岛隔离阀密封性定期试验的影响研究

核电机组核岛隔离阀密封性定期试验一般在机组启动的动态排气后进行,是确保隔离阀安全功能的重要试验。本文分析了核电机组安注管线主回路隔离阀密封性定期试验的原理和验收准则的设计方法,提出了动排气剩余空气体积对隔离阀密封性定期试验影响的分析方法。以RCP320VP为例,通过CFD建模计算,分析了剩余空气体积对定期试验结果的影响,即一定范围内剩余空气体积升高不会对定期试验结果带来影响,若一回路剩余空气体积提升过高(高于40标准立方米),在ASME标准允许的最大泄漏口径前提下,原来的定期试验验收准则将不再适用。本文的研究对于优化定期试验监督、提升机组核安全水平、提升一回路剩余空气体积标准值具有较大的参考价值。     

压水堆核电站一回路工况变化对主泵主要机械性能的影响

论述了大亚湾和岭澳1000MW压水堆核电站反应堆冷却剂回路(一回路)主要瞬态工况对反应堆冷却剂泵的主要机械性能参数的影响,为避免主泵受瞬态干扰,以及通过改变系统参数调整来改善主泵机械参数提供了理论依据。     

反应堆系统冷却剂泵流量特性计算模型

反应堆的发热是靠反应堆系统一回路冷却剂循环带出堆芯之外的。一般情况下，冷却剂的流动是靠冷却剂主循环泵（主泵）来唧送的，特殊情况下，也需要靠回路中冷却剂的自然循环流动来完成。不论是哪种情况，主泵的流量特性直接影响着反应堆的安全。本文根据主泵的四象限特性图提出了一种用于计算反应堆系统稳态和瞬态工况下主泵流量特性的计算方法。该方法便于使用，其计算结果与Ｒｅｌａｐ５／ＭＯＤ２的计算结果进行了比较，二者符合很好，证明本文的模型完全可用于反应堆系统的稳态设计和瞬态事故分析。     

CPR1000核电机组主泵动平衡的一次加准方法研究及应用

受核电站一回路复杂运行工况的影响,一回路冷却剂泵(下称"主泵")经常发生振动高的问题,现场动平衡是解决主泵振动问题的重要手段,也是影响和制约核电站机组启动上行的关键因素之一.通过对大量同类型主泵动平衡数据的总结,提出了主泵动平衡的一次加准方法.该方法给出了 CPR1000机组主泵动平衡中加重质量、加重角度和残余振动的计...     

CPR1000核电厂未能紧急停堆的预期瞬态保护信号及缓解系统改进

CPR1000核电厂发生丧失正常给水-未能紧急停堆的预期瞬态(LOFW-ATWS)时,若温度调节(R)棒组和功率调节(G)棒组的调节功能不能及时作用或丧失,存在一回路超压的风险。为降低瞬态过程中的一回路压力峰值,避免超压的风险,本文提出了瞬态过程中增设反应堆冷却剂泵停运的保护信号及缓解系统改进方案,并采用THEMIS程序进行改进方案的验证分析。结果表明,该改进方案可有效降低LOFW-ATWS事故下一回路压力峰值,消除一回路超压的风险。     

基于TRACE的大功率非能动核电厂自动泄压系统误启动事故计算

基于最佳估算程序TRACE,对大功率非能动核电厂冷却剂系统和非能动堆芯冷却系统进行了建模分析,得到了自动泄压系统(ADS)阀门误启动事故下,一回路压力、破口流量、一回路水装量等参数的瞬态变化,并以此为基准工况,根据电厂实践经验,选取泵延迟工况和阀门半开工况进行敏感性分析计算,将计算结果与基准工况进行了比较与分析。结果表明:虽然不同的工况可能造成一回路水装量低于基准工况,但最小的一回路水装量仍未低于限值,堆芯始终没有裸露,大功率非能动核电厂的非能动专设安全设施能有效对一回路进行冷却和降压,防止堆芯过热,验证了大功率非能动核电厂发生自动卸压系统误启动事故后的安全性。     

一回路氮气稳压系统瞬态模拟研究

基于氮气稳压基本原理,采用集总参数法开发了氮气稳压系统瞬态模拟程序,该模型突破了现有独立稳压器模型的局限,实现了一回路系统与氮气稳压器的直接耦合,并采用浮动式核电站氮气稳压系统试验数据对程序进行了验证。在此基础上,提出了一种基于敏感性分析的氮气稳压系统设计方法,与现有设计方法相比,该设计方法可以得到氮气稳压系统的优化配置方案,同时通过适配性设计,可以确保氮气稳压系统在启动过程中,压力不超过一回路系统温度压力限制曲线。      

射流装置启泵过程的瞬态特性

射流装置由射流泵和主泵组成,引入MRX(Marine Reactor X)压水堆一回路系统中,有助于提升反应堆的固有安全性。反应堆启泵过程中,流量急剧上升导致堆芯温度变化,影响堆芯运行安全。通过计算流体力学(Computational Fluid Dynamics,CFD)方法对引入射流装置MRX一回路10%满功率(Full Power,FP)、17.5%FP和25%FP堆芯功率下启泵进行三维瞬态模拟,分析MRX一回路中射流装置流场瞬态特性。结果表明,射流装置的加入可以改善一回路自然循环能力,提高启泵工况下冷却剂初始变化流量,减缓变化趋势,改善过渡安全性;启泵过程中一回路温度存在波动现象,且堆芯功率越大,波动幅度越大,时间越长;启泵完成后射流泵喷嘴处流速较大。验证了压水堆中引入射流装置提升反应堆固有安全性的可行性,同时为进一步优化设计方案提供方向参考。     

方家山核电厂主泵轴密封水系统运行分析

应用ANDRITZ冷却剂泵轴密封系统功能性试验中所得到的运行数据,分析轴封注入水密封的风险,主泵阀门状态改变对主泵运行的影响,主泵轴密封系统各级参数的变化对主泵启动及运行的影响。结果表明,轴封注入水密封存在一定的运行风险;主系统压力对轴密封注入水的高、低压泄漏流量影响不大;主系统的压力值应高于2.75 MPa时,主泵启动才是安全的。     

Two-phase flow in a full-scale loop seal facility

An experimental program has been carried out to study two-phase behaviour of a PWR cold leg loop seal during loss-of-coolant accidents. The experimental facility comprises a full-scale cold leg with a reactor coolant pump simulator. Three separate air/water test series were performed to determine the onset of slugging in the horizontal pipe, the residual water mass and the total two-phase pressure drop in the loop seal.The results of flow regime transition experiments have been compared with smaller-scale experiments and with theoretical predictions to evaluate scaling criteria. The strong hysteresis of transitions found between the stratified and slug flow regimes depends on the loop seal geometry and U-tube oscillations.     

Experimental study on the natural circulation of a passive residual heat removal system for an integral reactor following a safety related event

Natural circulation characteristics of an integral type reactor during the operation of a passive residual heat removal system (PRHRS) following a safety related event has been experimentally investigated by using the VISTA facility. A PRHRS actuation trip signal is generated by a high power trip signal following a steam flow increasing event. The experimental results show that the single-phase coolant flows steadily in the primary loop by a natural convection process and that it effectively removes the decay heat from the core through a steam generator during the PRHRS operation. The heat transfers through the PRHRS heat exchanger and the emergency cooldown tank (ECT) are sufficient enough to enable a two-phase natural circulation of the coolant in the PRHRS loop.     

Natural circulation experiments with a VVER reactor geometry

Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after Soviet-designed VVER pressurized water reactors. Using stepwise inventory reduction and small-break experiments, primary loop flow behaviour was studied over a range of coolant inventories. The tests revealed a trend toward decreasing primary side mass flow rate with inventory. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs and coolant flow into the hot legs changed from single to two-phase flow. The cause of this flow interruption was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. Finally, an experiment was conducted to demonstrate how loop seal refilling behaviour at low coolant inventories depends upon the steam flow rate through an individual hot leg. It was shown that loop seal refilling results when low steam velocities permit countercurrent flow in the upflow side of the loop seal.     

A new method to derive one set of scaling criteria for reactor natural circulation at single and two-phase conditions

Natural circulation is one of the most important thermal-hydraulic phenomena that makes the fluid flow along a closed loop without any external driving force. With this merit, it is adopted by the passive heat removal system to bring the residual heat out of the core at accidents, and by the primary system of some new conceptual reactors instead of pumps to drive the coolant in the loop at operation. To investigate the reactor natural circulation and verify system thermal-hydraulic codes, it is a way to construct an integrated effect test facility and perform experiments on it with the scaling criteria. With one-dimensional assumption, the natural circulation system was simplified as the heat source, heat sink and pipes, and described by two groups of equations independently for the single-phase and two-phase flow conditions. Based on these equations, a set of non-dimensional equations were derived and the criteria were obtained both applicable for single-phase and two-phase natural circulation. According to these criteria, the practical application was analyzed and discussed. In the paper, the property similarity was strongly suggested in most cases. Though equal height simulation was widely used in the past, the reduced height simulation is a good way to reproduce three-dimensional (3D) phenomena that are of concern in the investigation. The CHF simulation is not suggested. The mass of metal and its distribution is of concern instead of heat transfer at transient simulation.     

Stability of the water-chemistry regime in the loop of a passive safety system of nuclear power reactors

The organization of the water-chemistry regime in the loop of a passive safety system, whose purpose is emergency removal of heat from the core of a nuclear power reactor, is examined. It is shown that a selfregulated water-chemistry regime in which gaseous products of radiolysis can be dissolved in water coolant and recirculated into the irradiation zone, which will intensify liquid-phase radiation-chemical reactions of hydrogen with oxygen and organic release of gases from the liquid phase into the vapor-gas phase of the coolant, can arise in the loop of a passive safety system. This will result in the establishment in the loop of dynamic equilibrium between the release and dissolution of gases and will enable prolonged functioning of the safety system without intervention from the outside. The physicochemical and technical criteria for the appearance of a self-regulated water-chemistry regime for closed loops with natural circulation of the two-phase coolant are formulated and substantiated.     

Two-group interfacial area transport equations at bubbly-to-slug flow transition

In relation to nuclear reactor accident and safety studies, experiments on hot-leg U-bend two-phase natural circulation in a loop with a relatively large diameter pipe (10.2 cm inner diameter) were performed for understanding the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWRs. The loop design was based on the scaling criteria developed under this program and the loop was operated either in a natural circulation mode or in a forced circulation mode using nitrogen gas and water. Various tests were carried out to establish the basic mechanism of the flow termination as well as to obtain essential information on scale effects of various parameters such as the loop frictional resistance, thermal center and pipe diameter. The void distribution in a hot-leg, flow regime and natural circulation rate were measured in detail for various conditions. The termination of the natural circulation occurred when there was insufficient hydrostatic head in the downcomer side. The superficial gas velocity at the flow termination could be predicted well by the simple model derived from a force balance between the frictional pressure drop along the loop and the hydrostatic head difference. The bubbly-to-slug flow transition was found to be dependent on axial locations. It turned out that the formation of cap bubbles in the large diameter pipe caused the increased drift velocity, which would affect the prediction of the void fraction in the hot leg.     

Verification of the two-phase stratified-flow model in by separate effect tests

The code which is being developed by the Gesellschaft für Anlagen- und Rcaktorsicherheit (GRS) mbH is intended to cover, by means of a single code, the entire spectrum of loss-of-coolant and transient accidents in pressurized and boiling water reactors. The actual version Mod 1.1-Cycle A has a five-equation two-phase model based on the conservation laws for liquid mass, liquid energy, vapor energy and overall momentum. The relative velocity between liquid and vapor is determined by a full-range drift-flux model for two-phase flow in horizontal and vertical pipes. The verification of this drift-flux model is carried out by both large-scale experiments and single-effect tests. The single-effect test ECTHOR investigates stratified flow during the clearance of a water-filled loop seal by a forced air flow through the loop. ECTHOR is a French test for the consideration of two-phase flow regimes in pipes for the development of the codes. The experiments are dedicated to investigating typical two-phase flow during small break loss of coolant accidents (LOCA) in pressurized water reactors (PWR).As a measure, the remaining water level in the loop is determined as a function of the air flow rate. For the verification, a comparison between and computations, on the one hand, and experiments on the other hand is carried out. The results compare very well to each other. Test runs on different numerical grids show convergence to an asymptotic limit with increasing grid refinement.     

一回路相关死管段与热分层危害分析

为避免死管段与热分层危害，结合有关经验与核岛工艺系统设计特点，对某新型压水堆一回路各连接管逐一进行死管段与热分层危害分析。筛选出危害可能发生的管段后，对其中典型的热段连接余热导出管段应用计算流体力学软件CFX模拟分析，计算达收敛状态后可得出该管段热分层温度分布情况。另外，该管段下游两个隔离阀间封闭管段初始条件设定为充满工质，因受一回路影响而被加热升温，通过该封闭管段工质最终温度结果可判断是否出现死管段现象。最终计算数据显示热段连接余热导出管段总体上满足热分层验收准则，不过下游隔离阀间封闭管段有形成死管段的风险，但通过调整布置等措施可避免死管段危害。结果还显示出浮力循环流与一回路紊流冲击影响的流线特点。     

Effect of inlet geometry on hot-leg U-bend two-phase natural circulation in a loop with a large diameter pipe

In relation to nuclear reactor accident and safety studies, experiments on hot-leg U-bend two-phase natural circulation in a loop with a relatively large diameter pipe (10.2 cm ID) was performed for understanding the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR. The loop design was based on the scaling criteria developed under this program and a horizontal section was inserted between the gas injector and the hot leg in order to investigate the effect of the vapor phase inlet section on the flow regimes and flow interruption. The loop was operated either in a natural circulation mode or in a forced circulation mode using nitrogen gas and water. Various tests were carried out to establish the basic mechanism of the flow termination as well as to obtain essential information on scale effects of various parameters such as the loop frictional resistance, thermal center, and pipe diameter. The void distribution in a hot leg, flow regime and natural circulation rate were measured in detail for various conditions. The termination of the natural circulation occurred when there was insufficient hydrostatic head in the downcomer side. The superficial gas velocity at the flow termination could be predicted well by the simple model derived from a force balance between the frictional pressure drop along the loop and the hydrostatic head difference. The bubbly-to-slug flow transition was found to be dependent on axial locations. It turned out that the inlet geometry affected the flow regime at the inlet of the hot leg, namely the void distribution in the hot leg.     

气空间对自然循环两相流动系统稳定性影响实验研究

微沸腾工况运行是核供热堆实现热电联供的关键性问题之一，微沸腾运行工况下，两相流系统稳定性更加不利和复杂。通过实验研究，揭示了气空间对两相流系统稳定性的影响，研究提出通过气空间改性来抑制系统不稳定。实验结果表明，在气空间加装隔离孔板，对两相流系统不稳定振幅有明显的抑制作用，对两相流系统不稳定边界也有改善。 