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1.
J. Wilhelm 《Nuclear Engineering and Design》1987,103(1)
The development of filter systems for air cleaning in nuclear power plants will be briefly described. The result of research work on iodine filters was the basis for the use of gasketless deep-bed filters and of the multiway sorption filter in German reactor stations. The composition of the iodine release to the environment was validated with the “discriminating iodine species sampler”. The main sources for the release of elemental iodine from BWR and PWR were discovered and finally suppressed. The mechanical strength of HEPA filters is being tested at high temperatures and humidities. Prototype HEPA filters have been developed with much higher resistance against humidity, differential pressure, and corrosion. A filter for the removal of particles during extreme operating conditions such as containment venting was developed using stainless steel fibers for the filter medium. The first filter of this type has already been built and installed in a modern German PWR as a part of the containment-venting system for serious accidents. 相似文献
2.
D. Mitrakos S. Chatzidakis E.P. Hinis F. Parozzi 《Nuclear Engineering and Design》2008,238(12):3370-3378
In the course of a severe accident, some nuclear aerosols may be released to the environment through penetrating the containment concrete cracks, even if a catastrophic failure of the containment does not occur. There is experimental and theoretical evidence of strong retention of aerosol particles in the cracks that act as a filter. In this work a Eulerian model is developed based on the numerical solution of the one-dimensional aerosol transport equation. Plug formation is accommodated by allowing the crack diameter to change with time, based on the volume of the deposited mass. Brownian diffusion, gravitational settling and turbulence-driven deposition are considered as the removal mechanisms of the particles along the leak path. The model is verified against analytical solutions and validated by comparing with early as well as recent experimental data. It is concluded that a one-dimensional model of aerosol flow through a hydraulically equivalent leaking duct can simulate with enough accuracy aerosol transport in cracks, so that it may be an appropriate option for a large system code like ASTEC. 相似文献
3.
α-放射性气溶胶取样膜是放射性气溶胶监测系统的重要组成部分,选择过滤效率高、自吸收小及高表面收集特性的滤膜,将有利于提高监测速度和被测α的能量分辨率,后者有利于α辐射体的准确监测和成分分析。研究选用三种不同类型滤膜采集氡子体气溶胶,测定了滤膜对气溶胶取样的过滤效率、自吸收因子、表面收集特性以及阻力与流量的关系参数。在气溶胶浓度、取样流量和取样时间相同的条件下,平均孔径为0.8μm的混合纤维素酯滤膜的过滤效率最高,自吸收损失较小;孔径为0.4μm的重离子微孔滤膜的自吸收损失最小,表面收集特性优越,对于提高监测分辨率具有优势;玻璃纤维滤膜的阻力小,适合大流量取样条件下采用。 相似文献
4.
目前核电厂安全壳放射性评估中未考虑狭窄裂缝(简称窄缝)对气溶胶的滞留效果,但与常规尺寸相比,窄缝的高表面/体积比对气溶胶泄漏具有可观的滞留,评估结果过于保守。通过开展矩形直通道内气溶胶泄漏实验,获得缝高约100 μm钢制安全壳窄缝内气溶胶滞留效率,观察到窄缝通道入口区域为主要的粒子沉积区域。同时,通过在窄缝流动方向上建立并维持一定的温度梯度,模拟安全壳非能动冷却系统投运时安全壳窄缝内气溶胶泄漏过程。结果表明,窄缝对亚微米粒径气溶胶具有良好的滞留效果,温度梯度引入的蒸汽冷凝能显著提高气溶胶滞留效率至91%左右,且缩小了泄漏面积。 相似文献
5.
V.M. Mocho 《Nuclear Engineering and Design》2011,241(5):1785-1794
The IRSN and AREVA NC are currently conducting a common interest fire research programme with the aim, among other things, of improving knowledge of clogging of high efficiency particulate air (HEPA) filters and developing an empirical model for clogging of such filters by combustion aerosols. This model must - insofar as possible - be independent of the nature of the fuel and be able to be integrated in a calculation code covering the interaction between the ventilation and the fire. This paper discusses the influence of various “direct” factors such as the filtration velocity, the mass of deposited aerosol per filter area, the diameter and morphology of the combustion particles, the condensate content of the aerosols, and “indirect” factors such as the air flow feeding the fire and its oxygen content, which influence the evolution of the aeraulic resistance of a clogged filter. 相似文献
6.
为制备大流量气溶胶滤膜标准γ体源,采用参数优化的专用滴源装置,将放射性标准溶液60Co定量滴注在气溶胶取样滤膜上,晾干后对角折叠,并采用专用压样装置压制成与测量样品几何形状一致、密实的圆盘形固态物,密封保存。采用差重法对制备的滤膜标准γ源进行活度定值,制备的气溶胶滤膜标准γ体源活度值为4.4×104 Bq,相对扩展不确定度U(k=2)为3.6%。采用分割压制法对气溶胶滤膜标准γ体源进行均匀性测量,相对标准偏差低于5%。制备的滤膜标准γ源可为测量大流量气溶胶样品的γ谱仪效率刻度服务。 相似文献
7.
可溶性气溶胶(以下简称气溶胶)的吸湿增长是影响核反应堆严重事故中放射性产物动力学行为的关键因素之一,本文对吸湿增长的理论模型、实验测量方法以及核安全领域吸湿增长的研究进展进行了总结。核安全领域气溶胶吸湿增长理论模型以K?hler理论为基础,描述了吸湿增长过程中气溶胶热物性等参数与环境参数的关系,在此基础上发展的多种改进模型更适用于实际问题的分析,已在NAUA-HYGROS、MELCOR等核安全气溶胶计算程序中得到广泛应用。实验测量也是研究气溶胶吸湿增长特性的重要手段,相比于可测量气溶胶整体吸湿能力但结果较为粗糙的重量法,电力平衡法、HTDMA法精度较高,且具备实时测量单个颗粒及多模态颗粒群的吸湿能力,在核事故气溶胶吸湿增长特性实验研究方面具有潜在应用前景。本文最后总结了严重核事故领域吸湿增长的现有应用研究,包括核事故典型气溶胶吸湿增长特性的理论模型与数值计算应用研究、吸湿增长实验研究,数值计算研究表明,将气溶胶吸湿增长特性纳入到核事故气溶胶计算程序中可以实现对核事故发生后气溶胶行为更为准确的预测分析,相关实验得到了CsOH、CsI等典型核事故气溶胶的吸湿增长特性。 相似文献
8.
A steam generator tube rupture (SGTR) in a pressurized water reactor (PWR) might be a major source of accidental release of radioactive aerosols into the environment during severe accident due to its potential to by-pass the reactor containment. In the ARTIST program, tests were carried out at flow conditions typical to SGTR events to determine the retention of dry aerosol particles inside a steam generator tube. The experiments with TiO2 agglomerates showed that for high velocities in the range of 100-350 m/s, the average particle size at the outlet of the tube was significantly smaller than at the inlet due to particle de-agglomeration. Earlier, particle de-agglomeration has not been considered significant in nuclear reactor severe accidents. However, the tests in ARTIST program have shown that there is a possibility that TiO2 aerosol particles de-agglomerate inside a tube and in the expansion zone after the tube exit under SGTR conditions.In this investigation, we measured TiO2 aerosol de-agglomeration in the tube with very high flow velocities with two different TiO2 aerosols. The de-agglomeration was determined by measuring the size of the agglomerates at the inlet and outlet of the test section. The test section was composed of tubes with three different lengths, 0.20, 2.0 and 4.0 m, followed by an expansion zone.The main results were: (i) the de-agglomerate process was relatively insensitive to the initial particle size distribution, (ii) the agglomerates were observed to de-agglomerate in all the tubes, and the resulting particle size distributions were similar for both TiO2 aerosols, (iii) at high flow rates, increasing the gas mass flow rate did not produce further de-agglomeration, and (iv) the agglomerates did not de-agglomerate to primary particles. Instead, after de-agglomeration the particles had a median outer diameter Dc = 1.1 μm. Based on analysis using computational fluid dynamics (CFDs), the de-agglomeration was caused by the turbulent shear stresses due to the fluid velocity difference across the agglomerates in the viscous subrange of turbulence.It has to be noted that the particles used in this investigation were TiO2 agglomerates, and not prototypical nuclear aerosols with significantly different characteristics. Therefore, the results of this investigation cannot be directly used to determine whether the nuclear aerosol particles may de-agglomerate in SGTR sequences. However, this investigation highlights the possibility of particle de-agglomeration under SGTR conditions, and identifies the mechanism of the de-agglomeration inside the broken tube and when the aerosol is discharged to an open space. 相似文献
9.
Gyorgy Gyenes 《Nuclear Engineering and Design》2011,241(3):854-864
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. 相似文献
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11.
Aerosols generated by condensation of volatile fission products during nuclear reactor core meltdown accidents represent a major fraction of the accidental airborne radioactivity. A comprehensive experimental research programme was performed at Battelle to investigate the transport and deposition behaviour of aerosols in the containment, in order to support the development of computer models which estimate the fission product behaviour in the containment and the source term for potential radionuclide releases to the environment. Important steps in the investigations were: (1) DEMONA experiments. The first large scale aerosol test series performed in the Battelle model containment (BMC) (total volume 640 m3), using an open (quasi one-room) geometry and condensation aerosols from a plasma torch generator. (2) VANAM experiments. Advanced aerosol tests in the BMC, using a multi-room geometry, mixed hygroscopic/non hygroscopic condensation aerosols, a double injection period, and varying thermohydraulic conditions. One of the experiments was subject of the International Standard Problem ISP 37. (3) KAEVER experiments. A systematic investigation of aerosol materials and mixtures and the related deposition behaviour, using a simplified one-room test vessel (10 m3 volume) and advanced instrumentation. Important computer codes developed and/or validated in connection with the experiments are FIPLOC and NAUA; aerosol codes CONTAIN, MELCOR and GOTHIC-MAEROS were also applied. Some important results from the investigations and code developments are: (1) significant local aerosol concentration differences can occur in a multi-room geometry; (2) concentration differences can be caused by atmospheric stratification; and (3) deposition is strongly affected by material hygroscopicity and atmospheric humidity. (4) Satisfactory prediction requires a consistent treatment of multi-room thermal hydraulics, aerosol transport and steam condensation on particles. (5) Prediction results can be affected by numerical stability and nodalization (user experience). This paper presents a number of results of the experimental investigations and the present state of code modelling, with special reference to the findings of ISP37. 相似文献
12.
AP1000等非能动压水堆核电厂依靠自然的原理清除事故后安全壳气空间内的放射性气溶胶,可靠性较高,但对其进行分析较为复杂。事故后安全壳内气溶胶的主要运动形式有凝聚、重力沉降、扩散泳及热泳等,本文研究确定了合适的机理模型、假设条件和主要参数等,完成了AP1000核电厂的分析。分析结果表明,AP1000核电厂LOCA后,主要气溶胶去除机制中扩散泳贡献最大,其次是热泳和重力沉降;安全壳内气溶胶自然去除系数约为0.4~0.9h~(-1),堆芯裸露5h后变化较小;基于RG1.183源项、包络大气弥散因子及本文给出的安全壳气溶胶自然去除系数,计算得到的LOCA后厂外及主控室人员所受剂量可满足10CFR50中规定的限值要求。 相似文献
13.
《Annals of Nuclear Energy》2005,32(3):281-298
Containment structures not only provide a leak tight barrier, but also play a role in ensuring that the structures can withstand the impact load from projectile impacts or internal plant accidents. In assessing the containment structures of nuclear power plants, predicting the characteristics of impact resistance in relation to design and safety considerations is relevant. This investigation proposes a simple but effective method of performing numerical analysis on perforation resistance of reinforced concrete containment structures. In this work, normal and oblique impacting is considered to examine the residual velocity and impact phenomena of an ogive-nose steel projectile with various impact velocities against a reinforced concrete slab. Additionally, a phase diagram is devised to describe the ballistic terminal phenomena of projectile and target. This model could assess the resistance to penetration to results in the optimum design of the containment structures in nuclear power plants. 相似文献
14.
压水堆核电厂的高压熔堆事故覆盖了大部分的严重事故序列,且具有很大的潜在威胁。根据我国900MW压水堆核电厂的概率安全分析(PSA)结果选取了丧失厂外电、未能紧急停堆的预期瞬态、二回路管道破口、一回路小破口和蒸汽发生器传热管破裂5种典型的高压熔堆严重事故序列,并使用SCDAP/RELAP5程序对这些事故序列的进程和后果进行了计算分析。计算结果表明:5种典型高压熔堆事故序列可能导致高压熔喷和安全壳直接加热风险,可能引起安全壳早期失效,很有必要采取相应的一回路卸压措施。 相似文献
15.
Emmanuel Porcheron Pascal Lemaitre Denis Marchand William Plumecocq Amandine Nuboer Jacques Vendel 《Nuclear Engineering and Design》2010,240(2):336-343
TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear pressurized water reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety code validation. The present work is devoted to study a water spray injection used as a mitigation means in order to washout aerosol fission products. 相似文献
16.
为获取核材料化学爆炸事故烟云参数和气溶胶扩散规律,本文理论分析了核材料化学爆炸事故烟云扩散过程与机理,建立了基于CFD技术的烟云扩散数值仿真方法,并通过爆炸烟云外场扩散实验对该方法进行了验证。研究结果表明:CFD方法能实现烟云扩散的数值仿真;实验和仿真均显示小当量TNT爆炸高度远小于Church经验公式,宽度比例却增加;气溶胶颗粒在蘑菇云形态的双涡环烟云流场中分布不均匀,粒径大于50 μm的颗粒物大多位于烟柱中,而大部分可吸入颗粒物在烟云顶部聚集;气溶胶在烟云稳定前的驱散与沉降会改变其源项参数,以Operation Roller Coaster实验气溶胶积累质量分布为例的计算显示,空载释放率约为地面气化率的58%,可吸入比率由20%升高至31%,可吸入释放率约为18%。 相似文献
17.
CONTAIN-LMR是针对以液态钠为冷却剂的反应堆而开发的安全壳事故一体化分析程序。我国目前的CONTAIN-LMR程序版本为2000年左右从法国引进,还未进行过面向工程设计的系统性地程序开发和验证。本文主要针对CONTAIN-LMR程序中模拟池式钠火事故的分析模型进行详细分析,并采用国际上的池式钠火实验进行验证,实验验证结果表明CONTAIN-LMR程序可以较准确地模拟池式钠火事故造成的钠工艺间内的温度、压力升高及放射性钠气溶胶行为。本文的研究结果初步表明CONTAIN-LMR程序可用于钠冷快堆的钠火事故分析。 相似文献
18.
气溶胶样品氢含量的质子弹性散射分析 总被引:1,自引:0,他引:1
利用单个标准样品比较法质子弹性散射分析(PESA)对一组采集在核孔膜上的气溶胶样品的氢含量进行定量分析。在质子束能量为2.5MeV、散射角为40°时,PESA测量10μm厚核孔膜上气溶胶中H含量的探测限为0.36μg/cm2。几个样品H含量的测量结果在3.2~37.5μg/cm2之间,显著高于探测限。质子能量在1.0~2.5MeV范围内,3.5μmMylar膜的束流归一化H峰面积随入射质子能量的变化在±3%以内。所以,不同厚度的气溶胶样品可采用同一个标准样品来进行PESA测量。 相似文献
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Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code 总被引:1,自引:0,他引:1
Ivo Kljenak Maik Dapper Luis E. Herranz Joan Fontanet 《Nuclear Engineering and Design》2010,240(3):656-667
Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant. 相似文献