Analysis of source term aspects in the experiment Phebus FPT1 with the MELCOR and CFX codes

Several aspects related to the source term in the Phebus FPT1 experiment have been analyzed with the help of MELCOR 1.8.5 and CFX 5.7 codes. Integral aspects covering circuit thermalhydraulics, fission product and structural material release, vapours and aerosol retention in the circuit and containment were studied with MELCOR, and the strong and weak points after comparison to experimental results are stated. Then, sensitivity calculations dealing with chemical speciation upon release, vertical line aerosol deposition and steam generator aerosol deposition were performed. Finally, detailed calculations concerning aerosol deposition in the steam generator tube are presented. They were obtained by means of an in-house code application, named COCOA, as well as with CFX computational fluid dynamics code, in which several models for aerosol deposition were implemented and tested, while the models themselves are discussed.     

Aerosol retention in the flooded steam generator bundle during SGTR

A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out.To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated.Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with high gas flow rates carrying aerosol particles with them. However, compared to particle retention in the water close to the tube break, the effect of droplet entrainment on particle transport was small.     

压水堆释放源项快速估算程序开发

事故时向环境释放的源项是确定核电厂(NPP)应急响应水平和防护行动决策的重要依据。基于电厂工况估算源项是核电厂严重事故应急响应期间重要的应急评价内容之一。在国际原子能机构(IAEA)和美国核管会(NRC)的有关技术文档基础上,本文介绍了基于压水反应堆(PWR)工况进行事故释放源项估算的步骤和基础数据,并归纳了7种实用的事故释放源项估算方法。基于这些方法,开发了PWR事故时环境释放源项快速估算程序。该程序为不同估算方法提供4种释放途径:安全壳泄漏、安全壳旁通、蒸汽发生器传热管破裂(SGTR)和直接环境释放,除直接环境释放途径外,其他释放途径都估算了核素释放过程中的衰变、滞留、喷淋和过滤等减弱过程。对比发现,软件计算结果与美国核管会的RASCAL软件释放源项计算结果接近。     

ARTIST: introduction and first results

Aerosol Trapping In a Steam Generator (ARTIST) is a seven-phase international project (2003–2007) which investigates aerosol and droplet retention in a model steam generator under dry, wet and accident management conditions, respectively. The test section is comprised of a scaled steam generator tube bundle consisting of 270 tubes and three stages, one 1:1 separator unit, and one 1:1 dryer unit.As a prelude to the ARTIST project, four tests are conducted in the ARTIST bundle within the 5th EU FWP SGTR. These first tests address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities are sonic, and far away from the break, where the flow velocities are three orders of magnitude lower. With a dry bundle and the full flow representing the break stage conditions, there is strong evidence that the TiO₂ aerosols used (AMMD 2–4 μm, 32 nm primary particles) disintegrate into much smaller particles because of the sonic conditions at the break, hence promoting particle escape from the secondary and lowering the overall DF, which is found to be between 2.5 and 3. With a dry bundle and a small flow reproducing the far-field velocities, the overall bundle DF is of the order of 5, implying a DF of about 1.9 per stage.Extrapolating the results of the dry tests, it turns out that for steam generators with nine or more stages, it is expected that substantial DF’s could be achieved when the break is located near the tube sheet region. In addition, better decontamination is expected using more representative proxies of severe accident aerosols (sticky, multi-component particles), a topic which is yet to be investigated.When the bundle is flooded, the DF is between 45 and 5740, depending on the mass flow rate, the steam content, and the water submergence. The presence of steam in the carrier gas and subsequent condensation inside the broken tube causes aerosol deposition and blockages near the break, leading to an increase in the primary pressure. This has implications for real plant conditions, as aerosol deposits inside the broken tube will cause more flow to be diverted to the intact tubes, with a corresponding reduction in the source term to the secondary.     

CANDU steam generator life management

Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in PWRs. Canadian Deuterium Uranium (CANDU®) steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have resulted in a decrease in steam generator-related station unavailability of Canadian CANDU reactors. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development (R&D) work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for speciality tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service (FFS) guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. This paper will also show how recent advances in cleaning technology are integrated into a life management strategy. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New steam generator designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce-A/B, Pickering-A/B) and strategic plans to ensure that good future operation is ensured. The R&D program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factor.     

Ex-vessel releases: Aerosol source terms in reactor accidents

An ex-vessel aerosol and fission-product source term may arise from various events occurring in the containment building of a nuclear reactor. The research into the source terms associated with three of these events is reviewed. These source terms are from steam explosions, pressurized melt ejection, and melt/concrete interaction.The least is known about the steam explosion source term. Analyses indicate that its magnitude is likely lower than that assumed in the Reactor Safety Study (WASH-1400), but no conclusive experimental data are as yet available.The aerosol and fission-product source term from pressurized ejection of melt is an issue only recently addressed. Experimental evidence has allowed estimates to be made of the magnitude of this source term.The source term from melt/concrete interaction has been long recognized and has the largest data base. Experimental programs have addressed this source term for several years. A mechanistic model of material release has been developed and is discussed.     

Progress in understanding key aerosol issues

The 6th FWP SARNET project launched a set of studies to enhance understanding and predictability of relevant-risk scenarios where uncertainties related to aerosol phenomena were still significant: retention in complex structures, such as steam generator by-pass SGTR sequences or cracks in concrete walls of an over-pressurised containment, and primary circuit deposit remobilization, either as vapours (revaporisation) or aerosols (resuspension). This paper summarizes the major advances achieved.Progress has been made on aerosol scrubbing in complex structures. Models based on empirical data (ARISG) and improvements to previous codes (SPARC) have been proposed, respectively, for dry and wet aerosol retention, but, further development and validation remains, as was noted during the ARTIST international project and potential successors. New CFD models for particle-turbulence interactions have been developed based on random walk stochastic treatments and have shown promise in accurately describing particle deposition rates in complex geometries. Aerosol transport in containment concrete cracks is fairly well understood, with several models developed but validation was limited. Extension of such validation against prototypic data will be feasible through an ongoing joint experimental program in the CEA COLIMA facility under the 6th Framework PLINIUS platform.Primary deposit revaporisation has been experimentally demonstrated on samples from the Phebus-FP project. Data review has pinpointed variables affecting the process, particularly temperature. Available models have been satisfactorily used to interpret separate-effect tests, but performing integral experiments, where revaporisation is likely combined with other processes, still pose a difficult challenge. Further experimental data as well as modelling efforts seem to be necessary to get a full understanding. Resuspension, sometimes referred to as mechanical remobilization, has been recently addressed in SARNET and although a set of models were already available in the literature (i.e., Rock'n Roll model, CESAR, ECART), further work is needed to extend current capabilities to multi-layer deposits and to produce simplified, but sufficiently accurate, models. A major remaining uncertainty is the particle-to-particle/wall adhesion and its dependence on microscale roughness. Data from the previous EU STORM project have been retrieved and further experiments designed for code validation are being used to benchmark the models.     

Simulation of a gas jet entering the secondary side of a steam generator during a SGTR sequence: Validation of a FLUENT 6.2 Model

This paper summarizes the major insights gained as a result of gas jets entering a tube bundle from either a guillotine or a fish-mouth breach of a steam generator tube. This scenario is highly relevant in nuclear safety since it determines the potential retention of radioactive particles during risk-dominant sequences, the so-called Steam Generator Tube Rupture (SGTR) sequences. The scenario has been modeled with the FLUENT 6.2 code and its predictions have been proven to be grid independent and consistent with the experimental data available. The topology of the jets and the influence of the inlet mass flow rate (from 75 to 250 kg/h) have been studied in terms of velocity profiles.The results show that the breach shape heavily determines the jet topology. Both jets initially describe a quasi-parabolic trajectory, which is affected by the presence of the tubes. A guillotine breach generates a jet with azimuthal symmetry, which vanishes for the fish-mouth breach configuration. In this case, jet expands azimuthally in a pseudo-triangular way with a small angle. This fact diminishes the momentum loss across the bundle, so that for the same inlet mass flow rate the fish-mouth jet penetration is higher than the guillotine one. The normalized maximum radial and axial velocities of the jet from the guillotine breach are found to be self-similar with respect to inlet mass flow rate along the tube row position and axial distance to the breach, respectively. However, in absolute terms higher penetrations are found at higher mass flow rates.     

FRAMATOME's continous efforts to improve steam generator corrosion resistance

The steam generators of PWR nuclear reactors are among the primary components most affected by corrosion problems. Corrosion of the steam generator tubes, which assure heat transfer between the primary and secondary circuits, have been observed on a large number of operating steam generators, especially in the United States. According to an NRC survey, as of November 1981, forty PWR units with steam generators of the recirculation type were in operation in the US. Of these, 32 have been found to have one or more forms of tube degradation.Construction of the French PWR nuclear program started in the early 70s, at the time a number of operating plants in the US were being affected by the first corrosion problems. Since, at that time, its construction program was in an early stage, FRAMATOME was able to make modifications on the first units to improve steam generator resistance to corrosion. For instance, full depth expansion of the tubes in the tube-sheet using an explosive process (Westex) was performed on Fessenheim 1 steam generators already installed on site. Later on, continuous operating experience was being obtained in the US, before startup of the French units. This allowed FRAMATOME to react rapidly and take immediate corrective actions at the design stage, during fabrication and sometimes even on site in order to mitigate the risk of corrosion in the steam generators.FRAMTOME is confident that the present design of its steam generator models, including a large number of major improvements is adequate to prevent major corrosion problems to occur during operation. However, the company has embarked on an important development program to further improve the corrosion resistance and thereby the reliability of its steam generators. This program includes studies on new tube expansion techniques, alternate materials for steam generator tubes (Inconel 690), improved tube inspection methods, local thermohydraulic flow, tube vibrations, etc.     

核电厂卧式蒸汽发生器热工性能的计算和分析

文章基于卧式蒸汽发生器的工作原理及内部结构特点,建立了卧式蒸汽发生器数学物理模型,开发了针对卧式蒸汽发生器的热工水力程序。基于在役核电站卧式蒸汽发生器的设计参数,对程序进行了校核。该程序可以用来研究卧式蒸汽发生器内主要热工参数的分布情况,为卧式蒸汽发生器设计、安全分析提供指导;也可以根据在役核电站的历史运行数据对蒸汽发生器现阶段热性能进行分析评定,对蒸汽发生器一段时间内的热性能进行预测,为蒸汽发生器的运行、检修以及更换提供依据。     

Screening Test for Adsorbent for Kr at Liquid Argon Temperature under Nitrogen Atmosphere

A new acoustic leak detection system for sodium-cooled reactor steam generators using a delay-andsum beamformer is proposed. The major advantage of the delay-and-sum beamformer is that it could provide information on the acoustic source direction. An acoustic source of a sodium-water reaction is supposed to be localized, while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore, the delay-and-sum beamformer could distinguish the acoustic source of the sodium-water reaction from the steam generator background noise. In this paper, results of numerical analyses are provided to show the fundamental feasibility of the new method.     

De-agglomeration mechanisms of TiO2 aerosol agglomerates in PWR steam generator tube rupture conditions

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 D_c = 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.     

钢制安全壳窄缝内气溶胶冷凝滞留实验研究

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。      

核电厂蒸汽发生器与主泵对接焊缝检查系统开发

AP1000核电厂蒸汽发生器出口接管与主泵泵壳对接焊缝泵壳侧为粗晶奥氏体铸造材料,由于该焊缝壁厚大、超声衰减、晶粒散射严重等导致焊缝的超声检测技术开发难度大。本研究采用特殊的设计,开发了一套从蒸汽发生器出口接管内壁实施超声检测的自动检查系统,并将该系统应用于国内某AP1000核电厂的役前检查。结果表明,该检查系统完全满足现场检查要求,检验结果与焊缝出厂检验结果具有良好的一致性。      

Experiments and computational models for aerosol behaviour in the containment

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.     

An assessment of PWR steam generator condensation at the Oregon State University APEX facility

This report describes the testing to assess steam generator U-tube steam condensation conducted at the Oregon State University Advanced Plant Experiment Test Facility from 2005 to 2007. Six separate SG condensation (without non-condensable gas) tests were conducted as part of this test program. These tests were designed to evaluate steam condensation rates in a scaled Pressurized Water Reactor steam generator at various primary and secondary side pressures and inlet steam mass flow rates. The experimental data will provide a basis to assess TRACE steam generator modeling techniques and to assist in development of improved models for condensation and steam generator thermal-hydraulics.     

Anisotropic modelling of tube bundle flow during steam generator wet layup

Tube bundle flow can be considered as a porous medium flow and a fluid continuum can be established by introducing the porosity which is a ratio of fluid volume to total volume. Darcy's flow regime applies for the tube bundle flow of low Reynolds number during steam generator wet layup circulation. A general three-dimensional formulation appears as a steady-state heat conduction equation with source term and anisotropic conductivities. Solution to such an equation with appropriate boundary conditions can be obtained by any finite element computer program which solves anisotropic heat conduction problems. Capability of anisotropic modelling has been demonstrated by a sample problem of axisymmetric tube bundle flow with orthotropic hydraulic conductivities which are derived according to the existing empirical correlations for friction factors.     

后处理设施放射性液体贮罐泄漏事故环境释放源项估算

放射性液体泄漏事故是后处理设施典型的事故,泄漏事故通常发生在设备室。高放废液贮槽泄漏后气载放射性核素生成包括两个过程:一是在泄漏放射性液体的过程中惰性气体从溶液中释放,以及与空气、地板相互作用产生的气溶胶;二是泄漏后的蒸发过程(包括冲洗前稀释前和稀释后)。气溶胶在设备室内生成后会发生沉积,同时随着设备室排风系统,经过滤后向环境排放。本文给出了一种放射性溶液贮槽泄漏事故源项估算方法,实现了事故泄漏质量、泄漏活度、设备室气载放射性活度浓度及积分浓度、环境释放源项估算,为事故应急决策和响应行动提供数据支持。     

自然循环条件下蒸汽发生器倒U型管流量分配特性研究

以欧洲压水堆热工实验装置（PWR PACTEL）一回路系统蒸汽发生器为研究对象,首先,基于流体一维流动模型的质量、动量和能量守恒方程建立管道进出口压降以及传热与流体流量之间的关系;其次,以遗传算法为基础开发倒U型管蒸汽发生器流量分配计算程序,采用基准实验对程序正确性和可靠性开展验证;最后,利用流量分配程序计算蒸汽发生器倒U型管管组的流量分布情况,研究管高、管长以及一/二次侧换热系数对蒸汽发生器内流量分配的影响。结果表明,所开发流量分配程序计算结果与实验吻合良好;在选定的自然循环工况下,该蒸汽发生器中长管更易发生倒流,且倒流现象呈现分布范围广、单管流量低的特点;倒U型管内正流流速与管长成反比,与管高成正比,倒流流速随着管长的增加保持不变,与管高呈反比关系;传热系数较低时,总流量与传热系数成反比关系,当传热系数高于特定值后部分管内发生倒流,总流量骤降。      

基于多孔介质模型的快堆蒸汽发生器热工水力特性数值研究

蒸汽发生器（SG）作为钠冷快堆一次侧钠与二次侧水的热交换器,其可靠程度直接影响反应堆能否安全运行,因此对SG的一次侧热工水力特性的研究具有重要意义。本研究采用多孔介质模型,对快堆蒸汽发生器一次侧流场进行分析。通过对支撑板模型的计算,获得多孔介质控制方程的阻力源项。一次侧向二次侧的释热量通过系统程序Relap5计算,确定多孔介质控制方程的能量源项。通过用户自定义程序将动量源项与能量源项编译至FLUENT求解器中。通过FLUENT求解器求解控制方程,获得SG一次侧流场、压力场、温度场等信息。并通过对比模拟结果与设计值,验证了计算的准确性。