核电厂安全壳内气溶胶热泳沉积特性研究

核电厂事故下,裂变产物气溶胶沉积在热构件表面降低安全壳气空间内放射性。其中,由于构筑物、部件壁面温度梯度的存在,热泳沉积对气溶胶颗粒沉积的贡献不可忽略。本文采用符合安全壳气溶胶特性的公式计算了其在安全壳壁面的热泳沉积。结果表明热泳沉积效果随气溶胶粒径的增加而减弱;安全壳内壳表面温度梯度的提高,可以加强气溶胶的热泳沉积,从而提升安全壳内气溶胶的去除效果,降低安全壳内放射性水平。     

反应堆安全壳高精度气溶胶自然沉积模型的开发及验证

在压水堆严重事故过程中,气溶胶作为裂变产物的主要载体在安全壳内悬浮,有泄漏到外部环境中造成放射性污染的潜在威胁。安全壳气相悬浮的气溶胶会通过自然沉积机理沉降到壁面或地坑水,降低大气放射性。针对ISAA程序气溶胶模型精度不足的问题,改进安全壳气溶胶自然沉积模型。通过引入气溶胶动态形状因子,修正非球形气溶胶自然沉积速率,改进了重力沉积、布朗扩散、热泳和扩散泳沉积模型。选取AHMED(Aerosol and Heat Transfer Measurement Device)、ABCOVE(Aerosol Behavior Code Validation and Evaluation)和LACE(Light Water Reactor Aerosoal Containment Experiments)实验对改进代码进行评估。结果表明：改进模型能够更加精确模拟气溶胶质量峰值,响应安全壳压力温度对气溶胶自然沉积速率的影响,显著地提高了安全壳气溶胶残留质量的计算精度。改进后ISAA程序性能可以满足分析先进压水堆严重事故安全壳内气溶胶自然沉积行为的需要。     

微小通道内流动及气溶胶沉积特性实验研究

事故后气溶胶通过安全壳微小通道穿透会发生滞留现象，在不同载气流动状态下，主导沉积机制不同，会影响气溶胶的穿透率，因此需要研究微小通道内气溶胶的载气流动特性及气溶胶沉积特性。本文建立了微小通道气溶胶穿透与滞留实验装置，开展了雷诺数为40～3 600的载气流动与气溶胶沉积实验研究。在雷诺数约为700时，微小通道内会发生层流转捩。层流流动下，在大突缩比的矩形微通道中，随着雷诺数的增大，入口碰撞效应减弱，穿透率增大；小突缩比薄壁毛细管微通道入口效应弱，随着雷诺数的增大，气溶胶几乎完全穿透。在雷诺数为800～3 600的湍流流动下，气溶胶沉积主要处于涡流扩散-碰撞区及惯性主导区，受到湍流涡的影响气溶胶穿透率随雷诺数的增大而减小。     

压力驱动的毛细管内气体流动特性数值研究

安全壳缝隙内气溶胶的输运与滞留与缝隙内的气体流动密切相关。为利用现有的气溶胶穿透系数关系式评估事故后缝隙内的气溶胶滞留效率，以毛细管为代表性缝隙，基于一维可压缩绝热流动方程提出了一种压力驱动的气体流动计算方法，能够计算气流的亚音速、临界音速与壅塞状态。采用代表性的毛细管流动特性实验和计算流体力学(CFD)方法对本文提出的计算方法进行验证，结果显示流动变化趋势一致，气体流量与实验值相比绝大部分偏差小于10%,证明了一维计算方法的准确性与合理性，且效率优于CFD方法。在此基础上分析了毛细管两侧压差和内径对气体流动的影响，得到了不同内径毛细管内质量流量随压差的变化规律。本文研究结果可以为缝隙内气溶胶的滞留评估奠定技术基础，后续将拓展该计算方法至压力驱动的毛细管内多组分气体流动的模拟。     

微流体惯性撞击器气溶胶过滤行为研究

微流体惯性撞击器可作为安全壳过滤排放系统的第一级过滤装置，实现严重事故后安全壳快速泄压并对壳内放射性气溶胶进行初效过滤。为了探究惯性撞击器内气溶胶的过滤行为，开展可视化实验对气溶胶在惯性撞击器内的沉积分布进行研究。研究表明，可视化实验能反映撞击器内气流流向及气溶胶运动轨迹。气溶胶在过滤单元处的沉积位置主要集中在过滤单元的上表面及微流道的内壁面，过滤效率可达60%以上。同时，由于撞击器通道表面特性随气溶胶的沉积而改变，容尘条件下的气溶胶过滤效率逐渐增加。而当惯性撞击器内载气流速大于临界值时，过滤效率会因气溶胶的再悬浮而降低。     

压水堆严重事故下裂变产物迁移与释放研究

严重事故情况下大量放射性裂变产物释放到环境中,直接对人体造成危害。本文采用MELCOR程序,研究了DVI管小破口始发严重事故下,Cs I在一回路系统和安全壳中的存在形态,以及Cs I和惰性气体向安全壳、环境的迁移和释放行为。结果表明:Cs I主要以气溶胶形式释放出来,滞留在一回路系统中的Cs I几乎全部沉积在热构件上,约70%的Cs I以气溶胶的形式迁移到安全壳中,并不断沉积在安全壳热构件或溶于水中,极少量释放到环境中。事故后绝大部分惰性气体迁移到安全壳中,只有少部分滞留在反应堆其他系统中,在安全壳正常泄漏率下,释放到环境的惰性气体质量很少,仅为0.11%。     

非能动安全壳冷却对严重事故下气溶胶沉积影响分析

采用分区法模型计算了不同热工条件下的气溶胶沉积情况,分析了4种自然沉积机理对不同粒径气溶胶的沉积作用。研究表明,气溶胶扩散泳沉积受热工情况影响最为显著。针对AP系列压水堆非能动安全壳冷却的设计特性,可通过降低壁面温度来提高气溶胶的扩散泳沉积,增强安全壳内的气溶胶净化作用,从而提高严重事故下安全壳内的放射性去除效果。     

压水堆严重事故下气溶胶热泳沉积规律

通过编制程序,采用热泳沉积模型,计算严重事故工况下不同温度、压力、粒径时安全壳内气溶胶的热泳沉积效率。通过分析,可针对性地采取措施降低安全壳内壁面的温度,提高气溶胶的热泳沉积效率,增强反应堆的安全性能。      

环形窄缝通道内干涸后摩擦倍增因子的试验研究

在间隙为1.1和1.6mm的环形窄缝通道内,通过内外管双面通电加热流体,进行环形窄缝通道内的干涸后弥散流区压强的试验研究,通过试验得出环形窄缝通道内干涸后弥散流区全汽相摩擦倍增因子的计算式,该式能够较好地预测环形窄缝通道内干涸后弥散流区的两相摩擦压降。     

基于REVENT实验的气溶胶再夹带模型适用性分析

在核反应堆事故后卸压等特定场景下,安全壳内液体大量蒸发,液相中气溶胶在蒸汽作用下被夹带回气相中的现象称为再夹带。本文基于Revent实验结果对KataokaIshi's和Cosandey's再夹带模型的适用性进行了评估。首先将模型转化为程序语言,针对实验建立分析模型并对不同工况开展模拟研究;然后通过对比分析模型预测结果与实验测量结果,评估了在不同压力、气体组分条件下,KataokaIshi's各夹带区域模型预测可溶性气溶胶再夹带行为的适用性,Cosandey's模型预测可溶性、不可溶性气溶胶再夹带行为的适用性。结果表明:Cosandey's模型更适用于预测核电厂事故工况下安全壳内不同种类气溶胶粒子再夹带行为。     

Thermodynamic consequences of leakages in a double containment during severe accidents

For a large nuclear power plant under normal operating conditions a leakage rate for the containment of 0.25 vol.%/day is admissible. During a successfully controlled LOCA leakages of the containment will be released through filters by the annulus^* air exhausting system into the environment. During a core melt accident a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. When openings in the containment steel shell occur before a catastrophic failure, a depressurization into the annulus takes place. The area of the openings determines the depressurization rate and the thermodynamic conditions in the annulus. Furthermore the behaviour of the components being necessary for accident mitigation is influenced too. This paper discusses the thermodynamic consequences of leaks in the containment shell of a German PWR during a core melt accident. The results of those calculations are the necessary boundary condition for the estimation of fission product retention in the annulus.     

Research works on contamination transfers through cracked concrete walls

This study takes place within the framework of nuclear facilities containment assessment. The aims are to determine gaseous flow, two-phase flow and aerosol deposition models through a crack network by using the most realistic and representative crack network characteristics. For this, the crack network is considered as two infinite parallel plates. First of all, airflow experiments are performed on three concrete walls (128 cm in width, 75 cm in height and 10 cm in thickness), cracked by shear stresses. The results enable determining aeraulic crack network characteristics, thanks to the Poiseuille model in laminar compressible flow, and elaborating an experimental friction factor correlation for the transition flow, validated up to a Reynolds number of 250. Next, aerosol deposition experiments were performed with one of the previous concrete walls in order to determine a global aerosol deposition model in a crack network. The first experiments with an aerosol diameter of 60 nm showed that the aeraulic crack network characteristics are suitable for the aerosol physics. Therefore, geometrical crack network characteristics were determined by using several types of experiment and enable – thanks to aerosol deposition experiments – constituting and validating a global model of aerosol deposition in a crack network.     

T型三通管内热分层流动3D数值模拟

针对主管和支管流体温度不同的等径三通管,利用AQUA程序,对该三通管内的热分层流动进行了三维数值流场模拟。模拟结果显示,高流速的支管流体自上部射入,深入到主管的主流内部,迫使主管内上游来流贴着主管下壁面流动。在支管人流的背流面有一涡存在,主支流在此涡下部的狭窄通道内被加速到较支流速度更高的速度。对三通管内的温度分布计算表明,支流人流的迎流面存在着很大的温度梯度,两种流体的混合主要存在于远离支流人流的下游地区。     

Containment analysis on the PHEBUS FPT-0, FPT-1 and FPT-2 experiments

In a severe accident, most of the fission-product species are already condensed in aerosols when they are released to the containment. The behaviour of these aerosol particles controls the fission-product transport into the containment and affects the global Source Term. The calculations presented here were performed using the CPA module (Containment Package implemented in the European integral code ASTEC) for the in-pile PHEBUS FPT-0, FPT-1 and FPT-2 experiments and are focused on the aerosol transport. A detailed thermal-hydraulic model was used in the CPA/ASTEC code to evaluate the gas circulation pattern in the closed containment volume. The comparison of ASTEC results showed that the patterns are similar to the ones predicted by the CFD-based codes. Good agreement was reached with the measured average thermo-hydraulic parameters such as containment gas pressure, temperature and the condensation rate on the condensers. The calculations with the detailed simulation of the flow in the PHEBUS containment qualitatively predicted the particle settling on the elliptic bottom and deposition on the painted wet condenser surfaces. It was shown that the influence of the gas circulation leads to a relatively quick mixing of aerosols in the containment atmosphere. In the tests investigated, the effect of the gas circulation on the airborne aerosol mass during the aerosol injection period is small because the injected mass flux is significantly higher compared to the deposition fluxes on the vessel surfaces. During the long-term aerosol deposition phase, the flow fields predicted by CPA/ASTEC have a medium impact on the evolution of the airborne mass in the PHEBUS containment.     

The numerical implementation of multicomponent aerosol modelling

The assessment of the mitigation of the radiological consequences of a reactor accident by the retention of aerosol material in a containment building requires, in general, the separate modelling of the airborne mass and radioactivity. Such modelling has been implemented by generalizing the computer code aerosim. In simplified test cases the airborne activity is indeed found to deviate from the values which would be predicted assuming activity proportional to mass. The examples are also used to assess the numerical accuracy of the codes.     

Leak-tightness characteristics concerning the containment structures of the HTTR

The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area.The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values.The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.     

Air–steam leakage through cracks in concrete walls

In the context of a severe accident in a PWR nuclear plant, the evaluation of the leakage through the containment wall remains a key point of the safety analysis. Here we calculate the leakage of an air steam mixture through a traversing crack taking into account condensation. A 40 h test has been performed on a representative concrete slab with measurements of crack openings and flow rates. The CAST3M code enables us to simulate this test by making thermo-mechanical calculations and calculation of the leakage flow rate. Thermo-mechanical calculations provide data needed by the leakage calculations which are not measurable in the experiment. These are the internal crack profiles (variation of the opening with the curvilinear coordinate of the crack inside the concrete slab). Thermo-mechanical calculations are difficult to perform because boundary conditions of the test are complicated. Leakage calculations are performed with various hypotheses for the internal cracks profiles. A coefficient is applied on the friction factor to take into account additional complexity of the crack geometry.     

液池内气溶胶对孔板鼓泡体积影响的实验研究

液池内的孔板鼓泡是安全壳内气体过滤排放过程中的重要现象。过滤过程中,孔板鼓泡体积直接影响气泡的上升速度与气液接触面积,因此是影响过滤器过滤效率的重要参数之一。随着过滤的进行,液池内滞留的气溶胶可能成为孔板鼓泡体积的影响因素之一。本文采用可视化实验,对含BaSO₄和TiO₂气溶胶液池内的孔板鼓泡过程进行研究,观察和分析孔板鼓泡体积的变化规律,进而获取气溶胶对孔板鼓泡体积的影响机制。研究表明,高温液池和TiO₂会使得生成气泡体积增加,添加BaSO₄的影响并不明显,实验还发现了生成气泡顶部含“小气腔”的情况,表面张力及“小气腔”的变化是气泡体积改变的主要机制。      

Evaluation of plutonium oxide aerosol release from an LMFBR in a hypothetical accident

This work was undertaken to prepare a computer code for the hazard evaluation of plutonium oxide aerosol released to the atmosphere in the event of a hypothetical accident in a 50 MW(th) scale LMFBR. The reactor building structure consists of semi-double containments as follows: the primary containment has a large volume in comparison with the secondary annular containment in which a part is connected to the atmosphere through an emergency filter system. Sodium oxide aerosol containing PuO₂---UO₂ fuel, fission products and structural steel agglomerates quickly by coagulation due to its high concentration. Simultaneously, the aerosol concentration decreases due to settling, plating and thermophoresis. Using the present code, the amount of PuO₂ aerosol leakage to the atmosphere was evaluated.     

Experimental research on aerosols collection performance of self-priming venturi scrubber in FCVS

The self-priming venturi scrubber is the key component of filtered containment venting system to removal radioactive products during severe accident in nuclear power plant. In this paper, the collection performance of aerosols in the venturi scrubber is researched with experiment. The results indicate that the retention performance is closely related to the operating conditions and structures, and this relation is more closely for the removal of particle size under 0.5 μm. The retention efficiency in a venturi scrubber increase with improving of both gas velocity and injection flow rate, and the influence of gas velocity on efficiency is more effective at low injection flow rate. The venturi scrubber with a long throat length or small diffuser angle performs excellent retention performance for small size aerosols. In condition of gas velocity higher as U_avg = 200 m/s and the sufficient injection flow rate, the retention efficiency maintains upon 99% for the aerosol that size range from 0.1 μm to 10 μm. The pressure loss of the venturi scrubber increase slightly with extending the length of throat, and also with reducing diffuser angles. The removal efficiency is usually at the expense of the energy loss, while the higher aerosol retention efficiency corresponds to bigger pressure loss.