氚示踪剂在土壤包气层实验中的应用

测定了土壤包气层中氚在天然条件和人工喷淋条件下的扩散系数,研究了氚示踪剂在黄土包气层中的迁移规律。实验结果表明,在华北地区黄土高原天然条件下,包气带中氚示踪剂以０．０３ｃｍ／ｄ速度迁移,纵向扩散系数Ｄｘ为０．１０ｃｍ＾２／ｄ;在２ｍ＾２实验坑中,以５ｍｍ／ｈ喷淋强度每天喷淋３ｈ的条件下,氚示踪剂以２．２７ｃｍ／ｄ的速度向下迁移,Ｄｘ为７．７５ｃｍ＾２／ｄ。     

慢性γ射线照射复合苯,甲苯和一氧化碳吸入对动物细？…

四因素两剂量按正交表Ｌ８（２＾７）设计用家兔和小鼠在密闭舱内进行８组试验。γ射线的两剂量，对家兔为０．０３７５和０．００７５Ｇｙ／ｄ，对小鼠为０．０４７３和０．００９５Ｇｙ／ｄ，苯、甲苯和与经碳两浓度分别为１８２±３３和４０±１５ｍｇ／ｍ＾３、４０７±６８和９０±３０ｍｇ／ｍ＾３、２７８±８和９３±４ｍｇ／ｍ＾３。实验每天２ｈ，每周５ｄ，持续８周，结果如下：（１）γ射线、苯和甲苯对诱发家兔淋巴细胞     

低剂量X射线与化疗联合应用对小鼠移植肿瘤生长,转移及免疫功能的影响

将Ｌｅｗｉｓ肺癌细胞接种到Ｃ５７ＢＬ／６Ｊ小鼠右后肢腓肠肌内，当肿瘤直径达７ｍｍ时，予以不同治疗方案；联合治疗组；先给予７５ｍＧｙＸ射线一次全身照射，６ｈ后腹腔注射３．０ｍｇ／ｋｇＭＭＣ并建立肿瘤生长曲线，于治疗后１０ｄ计数肺转移瘤结节数，治疗后１２ｈ测定免疫功能。     

用角动量投影壳模型研究131La的转动带

将角动量投影壳模型应用到＾１３１Ｌａ核,对组态πｄ５／２,πｇ７／２,πｈ１１／２,πｈ１１／２⊙「ｖｈ１１／２」＾２、πｇ７／１⊙「ｖｈ１１／２」＾２的转动带理论计算和实验结果进行了比较,确定了各转动带的原子核形状。     

压水堆冷管段 2% 小破口失水事故实验研究

在高压综合实验装置（ＨＰＩＴＦ）上进行了压水堆冷管段２％小破口失水事故实验（ＮＳＢ－６），破口方向为冷管段底部，破口面积为２％。实验再现了核电厂发生小破口失水事故时的热工水力学现象，实验结果与ＲＥＬＡＰ５／ＭＯＤ２系统分析程序的计算结果作了比较，验证了该程序对小破口失水事故的分析能力。     

破口大小和二回路运行状态对核供热堆自然循环断流瞬态的影响

为研究一体化布置的核供热堆在发生破口失水事故中破口大小和从中间回路排出热量减少对断流过程的影响，选用不同的破口尺寸和不同的二回路工作状态，在５ＭＷ核供热堆热工水力模拟回路ＨＲＴＬ－５上进行了实验研究。稳态运行工况的系统压力为１．５ＭＰａ，在发生小破口失水事故后，加热功率维持为额定功率的５％以模拟剩余发热情况。实验研究并比较了不同条件下压力、温度、循环流量、液位和失水量等重要参数的变化。这些实验数据为核供热堆的安全分析提供了实验依据。     

PWR冷管段1%小破口失水事故实验研究

在高压综合实验装置（ＨＰＩＴＦ）上进行核电厂反应堆一次系统冷管段小破口失水事故（ＳＢＬＯＣＡ）模拟实验，破口方向为冷管段底部，破口面积为１％（ＮＳＢ－７工况）实验再现了核电厂发生小破口失水事故时的热工水力学现象，实验结果与ＲＥＬＡＰ５／ＭＯＤ２分析程序的计算结果上比较，验证了该程序对小破口失水事故的分析能力。     

~（99）Tc~m－d，l－HMPAO分解动力学及稳定方法的研究

报道了用国产药盒制备的￣（９９）Ｔｃ￣ｍ－ｄ，ｌ－ＨＭＰＡＯ（ＬＣ）分解动力学的研究。结果表明：ＬＣ分解反应为一级反应。在实验条件为２０℃、放射性浓度为１８５ＭＢｑ／ｍｌ、ｐＨ为７．０时，ＬＣ的分解速度常数ｋ为０．０１７±０．０７ｈ￣（－１）。ＬＣ溶液的稳定性与ｐＨ、放射性浓度及温度有关。在室温下，加入１ｍｍｏｌ／ｌ龙胆酸作稳定剂时，７４０ＭＢｑ／ｍｌ的ＬＣ能稳定３ｈ，放化纯度大于８０％。     

几种制剂对辐射所致脂质过氧化作用的影响

本文报道了γ射线照射对小鼠肝、脾、肾中脂质过氧化物（ＬＰ０）含量的影响，以及照后服用知母宁、单宁酸和鲨烯等制剂的抗脂质过氧化作用。制剂均灌胃给药，ＬＰＯ用改良的ＴＢＡ分光光度法测定。结果表明，γ射线整体照射小鼠后７２ｈ，其组织中ＬＰＯ含量增高，在０－４Ｇｙ范围内辐射剂量与ＬＰＯ含量呈线性关系；三种制剂均有降低小鼠肝、脾、肾中ＬＰＯ的作用，其中知母宁（０．３５ｍｇ／ｄ）的抗脂质过氧化作用显著优于单宁酸（１５ｍｇ／ｄ）和鲨烯（０．６ｍｇ／ｇ）。     

不同剂量 X射线全身照射对小鼠胸腺细胞周期进程的影响

本文采用细胞计量术研究了不同剂量Ｘ射线全身照射后小鼠胸腺细胞周期进程的变化。结果表明,７５ｍＧｙ Ｘ射线全身照射后２４～７２ｈ胸腺细胞周期各时相细胞数目发生明显变化,Ｇ０／Ｇ１期细胞数低于对照组,Ｓ期细胞数增多,说明其ＤＮＡ合成能力增强,至７ｄ时达到对照组水平;２．０Ｇｙ全身照射后４ｈ时胸腺细胞便开始出现明显的Ｇ２阻滞,１２ｈＧ２阻滞达最高点,７２ｈ其细胞周期进程明显加快,至７ｄ时达对照组水平。不     

Flooding in the pressurizer surge line of AP600 plant and analyses of APEX data

For the passive AP600 plant, the three stages of ADS (automatic depressurization system) valves are attached to the top of pressurizer. The existence of these valves makes liquid flow into and out of the pressurizer an important part of the dynamics during a small break loss-of-coolant accident. In this paper, counter-current flow limit (CCFL) in the surge line was analyzed. Specifically, CCFL in vertical piping, in slightly inclined horizontal piping, and in horizontal and vertical elbows were compared. The CCFL in the vertical section of the surge line was found to be the most limiting section. That is, the vertical CCFL controls the pressurizer liquid drain rate. This conclusion was tested and verified by comparing the predicted vertical CCFL against the counter-current flow states in the surge line, observed in small break LOCA tests conducted at the AP600 scaled test facility (APEX).     

过冷水流经通道的闪蒸临界流模型研究

反应堆发生失水事故时,破口处的临界流量决定着冷却水系统的装量,影响着堆芯燃料元件温度分布,对事故后果起重要作用。为了更好理解临界流动中各项参数的变化规律及机理,提出了两流体六方程临界流动模型,用来计算初始滞止状态为过冷水通过通道的临界流量。模型中既考虑了两相之间的动力学不均匀,也考虑了相间热力学不平衡。模型中引入了合适的计算闪蒸起始点位置和过热度的公式,并将汽泡增长方程与基本方程联立求解,可比较准确地反映汽泡的增长规律。在较宽的压力和温度范围内、不同长径比情况下,模型预测结果与试验结果符合较好,表明该模型具有较强的通用性。     

Two-phase countercurrent flow in a model of a pressurized water reactor hot leg

The onset of flooding or countercurrent flow limitation (CCFL) determines the maximum rate at which one phase can flow countercurrently to another phase. In the present study, the experimental data of the CCFL for gas and liquid in a horizontal pipe with a bend are investigated. The different mechanisms that lead to flooding and that are dependent on the liquid flow rate are observed. For low and intermediate liquid flow rates, the onset of flooding appears simultaneously with the slugging of unstable waves that are formed at the crest of the hydraulic jump. At low liquid flow rates, slugging appears close to the bend; at higher liquid flow rates, it appears far away from the bend, in the horizontal section. For high liquid flow rates, no hydraulic jump is observed, and flooding occurs as a result of slug formation at the end of the horizontal pipe. The effects of the inclination angle of the bends, the liquid inlet conditions and the length of the horizontal pipes are of significance for the onset of flooding. A mathematical model of Ardron and Banerjee is modified to predict the onset of flooding. Flooding curves calculated by this model are compared with present experimental data and those of other researchers. The predictions of the onset of flooding as a function of the length-to-diameter ratio are in reasonable agreement with the experimental data.     

Predicting local distributions of erosion–corrosion wear sites for the piping in the nuclear power plant using CFD models

The erosion–corrosion (E/C) wear is an essential degradation mechanism for the piping in the nuclear power plant, which results in the oxide mass loss from the inside of piping, the wall thinning, and even the pipe break. The pipe break induced by the E/C wear may cause costly plant repairs and personal injures. The measurement of pipe wall thickness is a useful tool for the power plant to prevent this incident. In this paper, CFD models are proposed to predict the local distributions of E/C wear sites, which include both the two-phase hydrodynamic model and the E/C models. The impacts of centrifugal and gravitational forces on the liquid droplet behaviors within the piping can be reasonably captured by the two-phase model. Coupled with these calculated flow characteristics, the E/C models can predicted the wear site distributions that show satisfactory agreement with the plant measurements. Therefore, the models proposed herein can assist in the pipe wall monitoring program for the nuclear power plant by way of concentrating the measuring point on the possible sites of severe E/C wear for the piping and reducing the measurement labor works.     

Off-take and Slug Transition at T-junction of Vertical-up Branch in the Horizontal Pipe

The off-take and the slug transition on air-water interface are experimentally investigated at the T-junction of the horizontal pipe with a vertical upward branch to simulate the loss-of-residual-heat-removal during a mid-loop operation in the Korea standard nuclear power plant. Scaling analysis is performed to scale down the experimental facility to the reference nuclear power plant. Two different diameters of branch pipes are used to verify the scaling laws and their scale effects. Air is used as working gaseous fluid and no water flow exists. Off-take behavior on horizontal stratified and slug flows is visually observed in the horizontal pipe. The experimental data are divided into three categories; onset of liquid entrainment at T-junction, onset of slug transition in the horizontal pipe, and discharge quality in the branch pipe. It is found out that the scale effect of the branch diameter on the onset of liquid entrainment is small and the existing correlations for it are applicable. Also, the onset of slug transition shows a discrepancy with Taitel-Dukler's correlation and has a strong influence on the discharge quality. New correlations for discharge quality are developed considering the critical dependency of the onset of slugging.     

Phase distribution in horizontal gas-liquid two-phase bubbly flow

1IntroductionTwo-phasebubblyflowisencounteredinawidevaxietyofindustrialapplications;Suchasintheproductionofelectricalpowerandpetrochendcajsindustry.Oneofthemosttwortantandyetleastunderstandingaspectsoftwo-phaseflowislateralphasedistribution.Theinformationonphasedistributionisarequiredparameterforhydrodynalincandthermalcalculationsinmanypracticalapplications.Considerableexperimentalandseal-theoreticalresearchhasbeenconductedbymanyresearchersinverticalupwaxdsordownwaxdsbubblynow.[1-7]However,L…     

Modeling of bubble shape in horizontal and inclined tubes

An experimental and theoretical study on the bubble shape of intermittent flow in the horizontal and inclined pipes has been carried out. The experiment results show that the bubble shape depends on the Froude number, bubble length and pipe inclination. The bubble with staircase pattern tail is observed at low Froude numbers, which is corresponding to plug flow. A model for the prediction of the bubble shape in horizontal and inclined pipes is proposed. The model is able to predict the bubble shape, flow pattern transition between plug and slug flow regimes as well as nose-tail inversion phenomenon observed in the downwardly inclined pipe. Validation shows the model can well predict the bubble shapes in horizontal and inclined pipes. The model discloses that the transition between plug and slug flow regimes occurs within a region. The Froude number range for plug flow regime in the downwardly inclined pipe is much wider than that in the horizontal or upwardly inclined pipe. The assumption of fully developed liquid film under the long bubbles tends to under-estimate the liquid fraction in this part of the slug structure, especially, for the intermittent flow in the upwardly inclined pipe with high Froude numbers.     

Investigation of liquid film behavior at the onset of flooding during adiabatic counter-current air–water two-phase flow in an inclined pipe

The liquid film characteristics at the onset of flooding in an inclined pipe (16 mm i.d. and 2.2 m in length) have been investigated experimentally. A constant electric current method and visual observation were utilized to elucidate the flow mechanisms at the onset of flooding. Two mechanisms are clarified to control the flooding in lower flooding and upper flooding, respectively. The lower flooding occurred at lower liquid flow rate and high pipe inclination angle. In this mechanism, the liquid film does not block the pipe cross-section. On the other hand, the upper flooding occurred at higher liquid flow rate and low pipe inclination angle. In this case, blocking of the pipe cross-section by large wave and entrainment plays an important role. The experimental data indicated that there was no reversal motion of liquid film at the onset of flooding during the operation of both lower flooding and upper flooding. The effects of pipe inclination angle on the onset of flooding are also discussed.     

Onset condition of gas and liquid entrainment at an inclined branch pipe on a horizontal header

An experimental and analytical investigation of onset condition of gas pull through and liquid entrainment in a single inclined branch to the arbitrary direction was performed in the present study. Since previous studies have been made only for specific angles of 0° and ±90°, it is hard to apply them to the case of branch off to the arbitrary direction. In the present study, series of experimental works was made with a horizontal header of 184 mm inner diameter with a single branch pipe which can be attached in nine different orientations of 0°, ±30°, ±45°, ±60°, ±90° angles and with two different diameters of 16.0 and 24.8 mm, respectively. Also, a simple mechanistic model is developed here based on the force balance at the top of the vortex on the surface of the stratified water. The onset condition due to the vortex breakup for the branch oriented with angle of θ was determined as and the onset condition due to the ascending film was which is similar to that of side branch. It was found that the present model predicted experimental data well in terms of the direction and diameter of the branch pipe. The present model and data may be useful in determination of the quality of flow through the inclined branch which is one of key parameter in the safety study of hypothetic small break loss of coolant accident in the nuclear power plant.     

Dynamic response of reactor internals to pipe breaks

This paper investigates the horizontal responses of the reactor internals to a 14 inch safety injection nozzle break, which is expected to cause the largest loads of the branch line pipe breaks defined for a nuclear power plant. It examines the effects of two forcing terms, reactor vessel motions and internal hydraulic loads, and suggests a new procedure which can be used for tributary pipe break analysis. The analysis result confirms the applicability of the suggested procedure to small size tributary pipe break analysis. Also, this paper calculates the horizontal responses of the reactor internals to a 3 inch pressurizer spray line nozzle break, which is the only one remaining on the primary side after leak-before-break evaluation, and secondary side pipe breaks such as main steam line and economizer feedwater line. The responses are compared with those of safe shutdown earthquake to show that seismic loads with a conservative margin may be used for the pipe break loads in the preliminary design.