反应堆压力容器下降段水-蒸汽CCFL实验与模型研究

为探究反应堆压力容器下降段在喷放末期冷段安注过程中的水-蒸汽逆流特性,建立下降段逆向流动限制(CCFL)模型,开展了基于压力容器模化本体的下降段CCFL实验研究以及建模分析。通过实验研究获得了不同入口安注水流量、安注水过冷度、堆芯蒸汽流量等条件下的下降段环腔内的安注特性数据,并基于实验数据进行了CCFL建模分析。结果表明,开始发生CCFL的蒸汽无量纲流速与入口安注水无量纲流速呈现正相关,基于无量纲流速建立的模型斜率与入口安注水无量纲流速呈现高度指数关联。本文建立了适用于从不发生CCFL至不完全CCFL,再到完全CCFL的下降段水-蒸汽气液逆流全过程预测模型。     

冷段大破口失水事故长期冷却及硼浓度特性研究

文章采用先进的热工水力分析程序CATHAR,对百万千瓦级ACP1000核电厂冷段大破口失水事故冷热段同时安注时CCFL作用下的上腔室及堆芯的流动换热特性、硼浓度特性进行了研究,并分析了破损环路热段安注流量大小对堆芯冷却的影响。研究表明：在热段安注总流量为614 m3/h时,破损环路对应热段安注流量的不同,不会对流入堆芯冷却有较大影响,破损环路热段安注流量差异不会对堆芯冷却有较大影响;切换至同时安注后堆芯硼浓度很快与系统达到平衡。     

LOCA下喷放参数对冷却剂喷放特性影响数值研究

冷却剂喷放过程是失水事故（LOCA）的重要过程之一,研究冷却剂喷放过程的热工水力特性对认识LOCA以及预测事故后放射性源项迁移过程有着重要意义。本文利用FLUNET软件建立冷却剂喷放数值计算模型,并对其进行验证。利用模型研究喷口直径、喷放距离和喷放压力等喷放参数对计算域内流场温度、液滴速度和蒸汽流速等特性的影响。研究结果表明：喷口直径的提高使得喷放参数均有提高;随喷放距离的增大,流场温度和液滴速度先上升后下降,而蒸汽流速先上升后趋于平稳;喷放压力越大,喷放参数的最大值离喷放出口越远,液滴速度和蒸汽流速的最大值随喷放压力的增大逐渐上升,而流场温度最大值没有变化。     

LOCA事故时一回路冷却剂管肘部回流流动极限研究

在压水堆发生LOCA事故时,需要依靠回流流动来进行堆芯的冷却,而存在着回流流动极限(Counter-Current Flow Limiting,CCFL),即冷却剂受重力作用向下流动时,会受到向上流动的蒸汽或其他气体阻挠,出现部分或全部冷却剂被气相带走的现象,导致冷却剂流速不能再增大,从而限制了传热效果。使用RELAP5对LOCA事故时弯头肘部的CCFL现象进行分析,分别研究管长L、管径D以及倾斜角θ对CCFL的影响,研究表明管长L越小,管径D越大,CCFL的安全裕度越高,而倾斜角θ对该条件下CCFL现象的影响不明显。     

WCOBRA／TRAC最佳估算方法在AP600大破口失水事故分析中的应用

AP600属于简易的先进压水堆设计，采用非能动安全系统，该系统起到了现行反应堆（McIntyre和Beck，1992）中能动的应急堆芯冷却系统（ECCS）的作用。为了验证AP600设计能够减缓假想大破口失水事故（LOCA）的后果，采用美国核管会最近批准的LOCA最佳估算方法（BELOCA），在标准安全分析报告中，对AP600大破口LOCA事故进行了分析。WCOBRA／TRAC程序针对AP600的特点进行建模，验证了匾柱型堆芯试验装置（CCrF）和上腔室试验装置（UPTF）的下降段注入试验的有效性，对AP600大LOCA事故工况下喷放和再淹没冷却传热的不确定性进行了再评估，保守的最小膜态沸腾温度用来定义喷放冷却的边界参数。由于采用了局部模型和总体模型，并采用了统计近似方法，再加上初始条件和边界条件的限制假设，BELOCA简化了对计算程序不确定性的定量计算。最终分析得到的95％包壳峰值温度（PCT95）为1186K，满足10CFR50．46的标准，并留有很大的裕量。本文因此得出结论：AP600的设计能够减缓假想大破口LOcA事故后果。     

600MW压水堆安注箱设计研究

本文用美国核管会热工水力程序TRACE和图形化建模软件SNAP,建立了600 MW两环路压水堆一回路和二回路热工水力系统分析模型,并对安注箱的各设计方案进行大破口失水事故(LBLOCA)模拟计算,通过对比各设计方案在LBLOCA事故下计算出的峰值包壳温度,研究安注箱在大破口失水事故工况下的安注性能,最后给出了优化的设计方案,并提出了可行的设计改进建议。研究结果表明,上腔室和下降段同时注入的方式较冷段注入和下降段注入更有效,且恰当地选取初始安注箱压力,可有效降低峰值包壳温度,提高LOCA裕量。     

反应堆下腔室涡流引起堆芯流动不稳定性试验研究

压水反应堆各个环路中的冷却剂在下腔室发生剧烈湍流交混,下腔室腔体内产生大量涡流,会导致堆芯燃料组件入口流量随机震荡,引发堆芯瞬态流动不稳定性,可能影响到反应堆热工、结构安全或传热性能。本文对反应堆内燃料组件区域流动特性开展研究,通过水力学试验手段获得反应堆堆芯在多种运行工况下,下腔室安装流量分配裙和不安装流量分配裙时的堆芯燃料组件入口流量脉动数据,试验结果表明,流量分配裙对下腔室涡流的抑制效果明显,在碎涡整流作用下,堆芯流量脉动明显降低;随着运行环路数的减少,下腔室流场对称性降低,涡流增强,堆芯流量脉动明显增大;下腔室涡流还会对堆芯入口流量分配均匀度造成不利影响,流量脉动偏大区域对应的流量分配因子明显较小。     

SuperLWR的LOCA分析

为了验证带有向下流水棒的超临界压力轻水冷却热中子反应堆（简称SuperLWR）的特性,进行了该堆的冷却剂丧失事故（LOCA）分析。分析范围为热／冷段的1％-100％破裂。在冷段大破口情况下,喷放期间过分的堆芯加热将通过自动卸压系统（ADS）得到缓解,这是因为反应堆卸压会导致堆芯冷却剂流动。在顶部水室和水棒内的冷却剂装量被有效地用于堆芯冷却。在喷放之后,像压水堆（PWR）一样,堆芯慢慢地被低压应急堆芯冷却系统（ECCS）再淹没。大破EILOCA的最高包壳温度低于准则值（1260℃）大约为430℃,出现在再淹没阶段。冷段的小破口给出了比大破口更高的包壳温度,这是因为在本分析中没有启动ADS。最高包壳温度低于准则值大约260℃。如果假定ADS被“干井压力高”信号启动,包壳温度将更低。热段破口不会比冷段破口严重,因为热段破口将增加堆芯冷却剂流量,预计在喷放之后将强迫淹没堆芯。     

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。     

含内热源有序饱和多孔介质通道内蒸汽-水两相流型可视化实验研究

采用均一球体形成多孔介质通道,通过高速摄像系统获得了多孔介质通道内两相流动影像数据,识别出多孔介质通道内蒸汽-水两相流动流型存在形式,并研究了部分参数对流型转变的影响规律。结果表明,多孔介质通道内的汽-液两相流型有泡状流、泡状-弹状混合流、弹状流、弹状-环状混合流以及环状流5种;随着入口过冷度的增加,泡状流向过渡流转变以及过渡流向环状流转变时所对应的汽相表观速度呈现出逐渐增大的趋势;随着压力的升高,泡状流向过渡流转变以及过渡流向环状流转变时所对应的汽相表观速度呈现出逐渐减小的趋势。      

Effect of Gas Atoms on Recovery of Work Hardening in Al-Li Alloys

In the first report of this study, dealing with CCFL and CCFL breakdown phenomena associated with the injection of emergency core cooling spray water into upper plenum during refill-reflood phase of a BWR LOCA, the following tests results were obtained.

The injected water maintained two-phase pool across the top of entire core after CCFL breakdown. The pool level oscillated near spray elevation. The objective of this paper is to clarify the mechanism of these phenomena, evaluating steam and spray flow effects on CCFL breakdown.

It is found that when spray flow rate was slightly larger than the CCFL drainage deter- mined by core steam flow, pool maintained at some constant level near spray elevation, after CCFL breakdown. On the other hand, when spray flow was appreciably larger than CCFL drainage, pool level slowly oscillated. The oscillation was caused by significant changes in steam condensation rate, and the corresponding subcooling penetration into the fuel bundles, when the pool level passed the spray elevation. The TRAC-BD1 analysis of test results suggested the small sector wall effect of test apparatus on CCFL breakdown phenomena.     

Analyses of subcooled CCFL tests for evaluation of W̱COBRA/TRAC applicability

The W̱COBRA/TRAC-MOD7A, Rev. 1 code is currently licensed for best estimate large break LOCA analyses of 3 and 4 loop PWRs with emergency core cooling system injection located in the cold legs. As a part of the licensing effort to extend the code applicability to an upper plenum injection plant, the codes ability to predict subcooled flooding on a perforated plate was assessed by analyses of GE counter current flow limit tests and by comparison to the Bankoff flooding correlation (Bankoff, S.G., Tankin, R.S., Yuen, M.C., Hsieh, C., 1981). Counter current flow of air/water and steam/water through a horizontal perforated plate, Int. J. Heat Mass Transfer, 24 (8) 1382). The observed code model bias for subcooled CCFL can be eliminated by applying multiplication factors to the interfacial condensation and the interfacial drag models.     

Condensation experiments for counter-current flow limitation in an inverted U-tube

In this study, we measured counter-current flow limitation (CCFL) characteristics in an inverted U-tube (18.4 mm diameter and 1.0 m straight-part length) simulating steam generator (SG) U-tubes under conditions of steam condensation at pressures of 0.1–0.14 MPa. Differential pressure ΔP between the top of the inverted U-tube and the lower tank was measured, and the flow patterns wave estimated by comparing the waveforms of ΔP with those in air–water experiments. As a result, we classified the flow patterns under CCFL conditions into CCFL-P, CCFL-L and CCFL-T. The falling water flow rate under CCFL conditions slightly increased as the pressure increased and the cooling water temperature decreased (subcooling of cooling water increased). In the case of CCFL-L, CCFL characteristics in the inverted U-tube were between those in air–water and saturated steam–water experiments at 0.1 MPa. Furthermore, we derived a Wallis type CCFL correlation and its uncertainty from CCFL data, including previously measured data, i.e., J*^1/2_G + 0.88J_L*^1/2 = 0.76 ± 0.05.     

T型管直接接触冷凝特性实验研究

在反应堆发生LOCA时,一回路系统压力降低,产生大量的蒸汽,安注水注入冷腿后可能会发生冷凝现象。为研究冷凝现象,通过开展T型管冷凝实验,在主管通纯蒸汽、支管通过冷水的情况下,研究了不同蒸汽流量和不同安注水流量下的冷凝量。结果表明:冷凝量存在一定的限制,即主管内蒸汽无法全部被冷凝。基于实验结果提出了一个冷凝效率与热力学比系数R_T之间的模型。     

Investigation of Break Location Effects on Thermal-Hydraulics during Intermediate Break Loss-of-Coolant Accident Experiments at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Break location effects on thermal-hydraulics during intermediate LOCAs were investigated by using four experiments at the ROSA-III, the 15 and 25% main recirculation pump suction line break (MRPS-B) experiments, the 21% single-ended jet pump drive line break (JPD-B) experiment and the 15% main steam line break (MSL-B) experiment. Water injection from the high pressure core spray (HPCS) was not used in any of the experiments. Failure of ECCS actuation by the high containment pressure was also assumed in the tests.

In the MRPS-B experiments, the discharge flow turned from low quality fluid to high quality fluid when the downcomer water level dropped to the main recirculation line outlet elevation, which suppressed coolant loss from the vessel and the core. In the JPD-B experiment, the jet pump drive nozzle was covered with low quality fluid and low quality fluid discharge continued even after the downcomer water level reached the jet pump suction elevation. Low quality fluid discharge ceased after the ADS actuation. It suggestes that the JPD-B LOCA has the possibility of causing larger and more severe core dryout and cladding temperature excursion than the MRPS-B LOCA. The MSL-B LOCA was characterized by mixture level swell in the downcomer and the core. The core mixture level swell resulted in the much later core dryout initiation than that in the MRPS-B LOCA, however, ECCS actuation was also delayed because of slow downcomer water level drop.     

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

Spray Cooling Test

A steady separate effects test on BWR spray cooling was performed at relatively high system pressures using the ROSA-III test vessel. These tests were conducted in order to promote a better understanding of the thermal-hydraulic phenomena in LOCA experiments and to obtain information necessary for improvement of analytical codes. The fraction of entrainment or overflow for various spray conditions was obtained and the data of CCFL at the upper tie-plate were compared with correlations. It was shown that the occurrence of CCFL significantly diminished core cooling effects and that rod quench by fall back water was quite irregular and unstable. Reflood core cooling was also studied.     

回流流动极限实验研究综述

回流冷凝是压水堆发生失水事故（ＬＯＣＡ）时的重要传热方式之一。当堆芯冷却剂严重丧失，自然循环又不能建立起来时，堆芯剩余释热依然能靠回冷凝带山。但在一定条件下会出现回流流动极限ＣＣＦＬ，并因此增加堆芯踝露程度。本文介绍了国外一些研究者在ＣＣＦＬ方面的研究成果，并建议开展ＣＣＦＬ研究。     

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.