共查询到18条相似文献,搜索用时 187 毫秒
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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。 相似文献
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为了研究小型反应堆在事故后亚微米气溶胶自然沉积行为,自主搭建了实验平台并开展了小冷凝速率下的相关实验。研究中发现蒸汽份额的提升对气溶胶基础的重力沉降过程存在促进作用,压力提升存在抑制作用;泳动去除机制的贡献占比随着蒸汽冷凝速率的提升而增加;冷凝速率较小时,热泳沉积机制在泳动去除机制中的占比可忽略不计;扩散泳S/W模型的适用性提高至385 K,当蒸汽密度和压力再增加时,实验所得亚微米气溶胶的扩散泳沉降速率高于S/W模型预测结果,根据蒸汽冷凝相关理论提出了修正系数。吸湿性气溶胶更容易在蒸汽冷凝条件下被扩散蒸汽夹带去除,3种扩散泳计算模型均无法准确预测吸湿性气溶胶的沉降过程。 相似文献
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严重事故时,安全壳内的多组分吸湿性气溶胶将在高湿度的条件下吸水增大,从而影响其重力沉降行为。通过理论分析,本文推导了多组分吸湿性气溶胶颗粒平衡粒径的物理模型,并通过实验结果进行验证。该模型重点关注溶解度对吸湿过程的影响,解释了多组分吸湿性颗粒粒径增大曲线不连续的原因。同时,分析了典型千兆瓦级压水堆核电厂中相对湿度、干粒径及非吸湿性组分质量分数对重力沉降去除系数的影响。结果表明,只有当气溶胶颗粒增大到一定程度后,其重力沉降速度才会明显的提高;对于干粒径超过0.01 μm的纯吸湿性气溶胶颗粒,只有超过一定湿度后其才会因吸湿而加速沉降,且该湿度下限随着干粒径的增大而减小;随着事故的进行,气溶胶颗粒中的非吸湿性组分质量分数逐渐增加,上述湿度下限将增加,且同湿度下吸湿对重力沉降的促进作用减弱。 相似文献
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研究应用GOTHIC8.0程序分析AP1000核电厂非能动安全壳冷却系统(PCS)传热传质过程,通过理论计算和程序分析两种方式对分析结果进行比较和评价。研究结果表明:GOTHIC8.0程序的DLM-FM模型适用于模拟安全壳内蒸汽在安全壳内壁面的冷凝传热传质过程,Film模型适用于模拟安全壳外水膜的蒸发传热传质过程。GOTHIC8.0程序可用于分析AP1000核电厂PCS传热传质过程,为AP1000核电厂在设计基准事故(DBA)下安全壳响应分析提供了另一种可行的工具。 相似文献
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以核电厂事故期间安全壳内放射性气溶胶通过缝隙向环境释放为研究背景,针对内径为397μm、长度为12 cm的毛细管,采用滑移通量模型对毛细管内气溶胶输运与滞留进行数值模拟研究。首先基于鹿儿岛大学开展的毛细管流动实验对本文建立的毛细管CFD模型进行了验证,结果表明CFD模型可准确模拟毛细管内压力驱动的一维可压缩绝热流动。在此基础上开展不同滞止压力下的气溶胶输运与滞留计算,结果显示在本文所考虑的重力沉降、布朗扩散和湍流扩散3种气溶胶沉积机理中,湍流扩散占主导作用;气溶胶穿透系数随滞止压力提高而下降,当滞止压力超过550 kPa后,穿透系数小于0.02。本研究为缝隙内气溶胶输运与滞留分析提供了新的技术方法,后续将在模型中增加其他沉积机理,并开展实验进行验证。 相似文献
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张琨 《核标准计量与质量》2012,(3):16-21
安全壳直接加热(DCH)是压水堆核电厂严重事故中的主要现象之一,可能导致安全壳早期失效、大量放射性释放的严重后果.国际上的核安全管理机构均非常重视该现象并制定了相关的法规要求.文章一方面概述与DCH相关的法规要求,另一方面针对AP1000核电厂发生DCH的事故工况,进行后果分析方法的研究.分析结果表明,AP1000核电厂的DCH不会造成安全壳失效. 相似文献
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根据AST方法建立了AP1000LOCA放射性核素活度计算模型,研究事故后安全壳及环境中放射性核素活度的变化。结果表明:事故后安全壳气空间内各核素的放射性活度呈先增大后减小的趋势,40min时达到最大。根据核素性质,将其分为不考虑母核衰变的核素和考虑母核衰变的核素。事故发生40min后,前者在安全壳内的活度指数减小,典型核素有131~135I、83 Krm等,后者由于母核衰变的影响导致其在安全壳内的活度减小趋势放缓,典型核素有85 Kr、133 Xem、133 Xe和135 Xe等。I和Cs由于受自然去除机制的去除作用,事故几小时后其向环境的累积释放量增长非常缓慢;对于Kr和Xe,半衰期较长的核素向环境的累积释放量不断增大,半衰期较短的核素在事故几小时后向环境的累积释放量趋于平衡。 相似文献
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反应堆发生严重事故时,堆芯释放出的吸湿性气溶胶会在潮湿的安全壳内增大,从而影响其自然去除过程。本文理论推导了吸湿性气溶胶的增大模型并通过多种方法对其进行了验证。模型计算结果表明,气溶胶的增大过程由于受到溶解度的限制而存在临界湿度值,在该临界值以下时气溶胶不发生吸湿,但这未被其他严重事故分析程序所考虑。同时,基于某三代先进压水堆的特定严重事故工况,本文分析了干颗粒半径及湿度对气溶胶的平衡半径和自然去除系数的影响。结果表明:气溶胶的自然去除系数随干颗粒半径的增大将先减小后增加,并在1 μm时达到最小值;相同湿度下,干颗粒半径对气溶胶半径的最大增大比例的影响不大;湿度的增加对不同干颗粒半径气溶胶去除系数的影响不同。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(11):823-837
A computer code MIRA-PB for predicting the iodine removal by containment spray in LOCA was prepared on the basis of MIRA-P/MIRA-B code developed in Battelle Columbus Laboratories. MIRA-PB considers behavior of inorganic iodine, organic iodide, and iodic aerosol and simultaneous removal by natural deposition, liquid-film absorption, spray washout, filtration and leakage to the environment. The iodine removal by the containment spray systems in LOCA of PWR and BWR is calculated with the MIRA-PB. 相似文献
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Some kinds of break in the reactor coolant system may cause the coolant to exit rapidly from the failure site,which leads to the loss of coolant accident (LOCA).In this paper,a stress analysis of an AP1000 reactor containment is performed in an LOCA,with the passive containment cooling system (PCCS) being available and not available for cooling the wall's containment.The variations in the mechanical properties of the wall's containment,including elastic modulus,strength,and stress,are analyzed using the ABAQUS code.A general two-phase model is applied for modeling thermal-hydraulic behavior inside the containment.Obtained pressure and temperature from thermal-hydraulic models are considered as boundary conditions of the ABAQUS code to obtain distributions of temperature and stress across steel shell of the containment in the accident.The results indicate that if the PCCS fails,the peak pressure inside the containment exceeds the design value.However,the stress would still be lower than the yield stress value,and no risk would threaten the integrity of the containment. 相似文献
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Advanced nuclear water reactors rely on containment behaviour in realization of some of their passive safety functions. Steam condensation on containment walls, where non-condensable gas effects are significant, is an important feature of the new passive containment concepts, like the AP600/1000 ones.In this work the international reactor innovative and secure (IRIS) was taken as reference, and the relevant condensation phenomena involved within its containment were investigated with different computational tools. In particular, IRIS containment response to a small break LOCA (SBLOCA) was calculated with GOTHIC and RELAP5 codes. A simplified model of IRIS containment drywell was implemented with RELAP5 according to a sliced approach, based on the two-pipe-with-junction concept, while it was addressed with GOTHIC using several modelling options, regarding both heat transfer correlations and volume and thermal structure nodalization. The influence on containment behaviour prediction was investigated in terms of drywell temperature and pressure response, heat transfer coefficient (HTC) and steam volume fraction distribution, and internal recirculating mass flow rate. The objective of the paper is to preliminarily compare the capability of the two codes in modelling of the same postulated accident, thus to check the results obtained with RELAP5, when applied in a situation not covered by its validation matrix (comprising SBLOCA and to some extent LBLOCA transients, but not explicitly the modelling of large dry containment volumes).The option to include or not droplets in fluid mass flow discharged to the containment was the most influencing parameter for GOTHIC simulations. Despite some drawbacks, due, e.g. to a marked overestimation of internal natural recirculation, RELAP5 confirmed its capability to satisfactorily model the basic processes in IRIS containment following SBLOCA. 相似文献
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M. Taube M. Lanfranchi Th. von Weissenfluh J. Ligou G. Yadigaroglu P. Taube 《Annals of Nuclear Energy》1986,13(12):641-648
A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydrodynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h. 相似文献