带定位格架的12根束棒内水表面沸腾时的放热系数实验研究

通过实验,证明了12根束棒带波纹片定位格架区水表面沸腾放热系数比稳定段低。提出了12根束棒稳定段水表面沸腾平均放热系数关系式。     

HWRR—3功率起伏分析

一、前言HWRR-3堆芯装载着72束3%浓缩度的UO_2燃料组件。每束组件内有12根圆形排列的燃料元件棒。活性区中央有一个大的水腔,等效直径31.15cm,内装有五根实验管道。     

稠密带缠绕丝棒束阻力特性研究

为获得稠密布置燃料组件的阻力系数,应用稠密带缠绕丝棒束进行实验研究,拟合阻力系数关系式,并将关系式与经典Rehme关系式进行比较分析。结果表明Rehme关系式不适用于本实验棒束。同时应用计算流体力学(CFD)方法、剪切应力输运模型(SST)湍流模型对实验进行模拟,获得棒束内部的流动形式、压力场和沿程阻力系数,并与实验结果进行对比。结果表明CFD方法可作为预测稠密带缠绕丝棒束单相流动阻力系数的参考。     

快堆带绕丝棒束组件低雷诺数下的水力特性分析

低雷诺数(Re)流动存在于正常运行或事故停堆工况的各类组件中,对于快堆的安全运行具有重要意义。利用CFX程序对低Re下的中国实验快堆不同类型的带绕丝棒束组件的水力特性进行了分析。结果表明,通过利用1个螺距的带绕丝棒束组件计算得到的低Re下的水力特性与实验结果以及Engel关系式符合较好。通过利用4个螺距的带绕丝棒束组件计算结果表明,绕丝产生的横向流动使组件6个壁面上压力分布有所不同,但在流动充分发展时,每个面轴线方向的压降按螺距均匀分布,从而进行带绕丝棒束组件水力特性测量时,需在组件同一面上按照整数倍螺距来布置测点,才能避免由于横向流动对测量带来的影响。     

快堆带绕丝棒束组件低雷诺数下的水力特性分析

低雷诺数(Re)流动存在于正常运行或事故停堆工况的各类组件中,对于快堆的安全运行具有重要意义。利用CFX程序对低Re下的中国实验快堆不同类型的带绕丝棒束组件的水力特性进行了分析。结果表明,通过利用1个螺距的带绕丝棒束组件计算得到的低Re下的水力特性与实验结果以及Engel关系式符合较好。通过利用4个螺距的带绕丝棒束组件计算结果表明,绕丝产生的横向流动使组件6个壁面上压力分布有所不同,但在流动充分发展时,每个面轴线方向的压降按螺距均匀分布,从而进行带绕丝棒束组件水力特性测量时,需在组件同一面上按照整数倍螺距来布置测点,才能避免由于横向流动对测量带来的影响。     

使用离散钍铀燃料组件的CANDU6堆物理特性初步研究

为研究钍铀燃料在CANDU6堆中的应用,采用DRAGON/DONJON程序,对使用离散型钍铀燃料37棒束组件的CANDU6堆进行时均堆芯分析。结果表明,组件采用235U富集度为2.5%的铀棒以及第1、2、3圈布置钍棒的37棒束组件,堆芯在8棒束换料、3个燃耗分区的方案下,组件的冷却剂空泡反应性较使用天然铀的37棒束组件（NU-37组件）与采用混合钍铀元件棒的37棒束组件更负;堆芯最大时均通道/棒束功率满足小于6700?kW/860?kW的限值;燃料转化能力比采用NU-37组件时更高;卸料燃耗可到达13400?MW·d/t(U)。研究表明,所设计的离散型钍铀燃料37棒束组件可用于现有CANDU6堆芯,且无需对堆芯结构及控制机构作重大改造;燃料组件和堆芯设计方案可为钍铀燃料在CANDU6堆芯的应用提供参考。      

六边形排列的7棒束通道内液态钠流动换热特性试验研究

以液态钠作为试验工质,对六边形排列的7棒束通道内液态钠流动换热特性进行了试验研究。试验流速为0~4 m·s^-1,热流密度为0~120 kW·m^-2,系统压力为1.5~200 kPa,对应的雷诺数和佩克莱数分别为4 000~60 000和0~340。深入分析了部分热工参数对7棒束通道内液态钠流动换热特性的影响,通过对7棒束通道内液态钠流动换热的试验数据的非线性拟合,得到适用于7棒束通道内液态钠流动换热的经验关系式。结果表明：拟合得到的摩擦系数关系式和努塞尔数关系式能准确地预测7棒束通道内的试验数据,其预测误差分别小于5%和6%。将获得的努塞尔数关系式与其他研究者的试验数据进行比较,与其他研究者985%的试验数据误差在30%以内,表明获得的关系式适用于7棒束通道内液态钠流动换热。     

螺旋管中汽-水两相流强制对流沸腾传热研究

对螺旋管中两相强制对流沸腾传热特性进行了试验研究。试验是在高压水回路上进行的 ,试验参数为系统压力 6 0～ 1 1MPa ,质量流速 40 0～ 1 2 0 0kg/(m2 ·s) ,热流密度 0～ 45 0kW /m2 ,螺旋直径1 3 7m ,螺旋上升角 3 94°。用修正L M关系式整理了两相强制对流放热系数。同时也得到了螺旋管单相水和单相蒸汽的强制对流放热系数 ,并与文献进行了比较     

紧密排列棒束燃料组件临界热流密度实验研究

基于高转换比紧密布置堆芯研究背景,针对堆芯紧密排列螺旋绕肋棒束组件开展了临界热流密度(CHF)实验研究,获得了棒束在不同热工条件下临界热流密度。研究结果表明:紧密排列棒束燃料组件CHF主要发生在热棒元件,临界发生时加热元件壁面温度迅速升高,同时压力升高,流量降低;系统压力、质量流速、含汽率、入口过冷度等热工参数对组件临界热流密度影响较大;获得了CHF计算关系式,计算值与实验值偏差在±10%以内。     

燃料棒束换热(RBHT)试验对低压下堆芯漂移流模型的评价

我国目前正在发展基于非能动技术的三代核电,为评价和改进非能动核电厂小破口失水事故在低压下棒束区的漂移流模型,采用燃料棒束换热(RBHT)试验对EPRI[6]、Cunningham-Yeh[4]模型,Bestion[7]漂移流模型进行了计算分析,结果表明燃料棒束换热试验RBHT试验数据工况能涵盖非能动核电厂在低压下的参数,不需要建造针对燃料棒束的试验台架,Cunningham-Yeh[4]和Bestion[7]漂移流模型基本适用,而EPRI[6]在低压区过高预测了空泡份额,不适用非能动核电厂。     

离子径迹结构对SRAM单粒子翻转截面的影响

本文基于北京HI-13串列加速器的单粒子效应测试终端对0.15 μm工艺的SRAM进行了单粒子效应测试,再次验证了在截面曲线接近饱和区部分,高能离子翻转截面低于低能离子翻转截面的现象。采用Geant4对其进行模拟研究,结果表明,相同LET条件下高能离子可在较远处沉积能量,更易使同一存储单元内相邻的节点共享电荷发生单粒子翻转恢复而减小其单粒子翻转截面,而低能离子进行单粒子效应测试的结果相对保守。     

Characteristics of the local bubble parameters of a subcooled boiling flow in an annulus

An experimental study on the subcooled boiling phenomena was carried out in the SUBO (SUbcooled BOiling) test facility under steam-water flow condition. The test section is a vertical annulus of which axial length is 4.165 m with a heater rod at the center of a channel. The inner and outer diameters of the test section and the heater rod are 35.5 mm and 9.98 mm, respectively. For the measurement of the local bubble parameters, double sensor optical fiber probes were applied at six elevations along the test channel. Among them, one is installed in the unheated region which is located downstream of the heated section for the measurement of bubble condensation. A total of six test cases was chosen for the parametric study of the heat flux of 370-563 kW/m², mass flux of 1110-2100 kg/(m² s) and inlet subcooling of 19-31 K at pressure condition of 0.15-0.2 MPa. From the test, local void fraction, interfacial area concentration, Sauter mean diameter and bubble velocity were measured at 11 radial locations at each elevation. The measured data shows well development and propagation of the bubble parameters along the test channel. The present data is expected to be suitable for a benchmark, validation and model development of the CFD codes or existing safety analysis codes.     

Modeling the heavy ion upset cross section

The standard Rectangular Parallelepiped (RPP) construct is used to derive a closed form expression for, σ¯(&thetas;, φ, L) the directional-spectral heavy ion upset cross section. This is an expected value model obtained by integrating the point-value cross section model, σ(&thetas;,φ,L), also developed here, with the Weibull density function, f(E), assumed to govern the stochastic behavior of the upset threshold energy, E. A comparison of σ¯(&thetas;,φ,L) with experimental data show good agreement, lending strong credibility to the hypothesis that E-randomness is responsible for the shape of the upset cross section curve. The expected value model is used as the basis for a new, rigorous mathematical formulation of the effective cross section concept. The generalized formulation unifies previous corrections to the inverse cosine scaling, collapsing to Petersen's correction, [cos&thetas;-(h/l)sin&thetas;]^-1, near threshold and Sexton's, [cos&thetas;+(h/l) sin &thetas;]^-1, near saturation. The expected value cross section model therefore has useful applications in both upset rate prediction and test data analysis     

Separate effects study on sodium boiling and dryout under forced-and natural-convection conditions performed with a one-dimensional LOOP code

A comprehensive separate effects study has been performed with the one-dimensional code LOOP-1 on powers and times for sodium boiling initiation and dryout in a closed loop system. Two different kinds of transients were considered: loss-of-flow and loss-of-heat-sink. Loss-of-flow transients were studied under both forced- and natural-convection conditions. Loss-of-heat-sink transients were studied under natural-convection conditions. The results for loss-of-flow transients indicate that the boiling initiation time was reduced by a small amount, and the dryout time was reduced very significantly by increasing either the input power or the inlet temperature, or by decreasing the test section pressure for both forced- and natural-convection conditions. Under forced-convection conditions, a stabilizing effect occured by either increasing the test section valve setting or by decreasing the bypass ratio with a pump head adjusted to provide the same steady state and initial transient flows; thus, longer boiling times could be maintained before dryout occurred. For natural-convection loss-of-flow conditions, increasing the test section valve setting or decreasing the bypass ratio reduced the test section inlet flow, which resulted in boiling inception and dryout occurring more rapidly. A larger flow before the loss of flow transient starts yielded longer boiling initiation and dryout times. Under loss-of-heat-sink conditions, the higher the inlet temperature, the lower the boiling and the dryout powers. The margin between boiling and dryout powers increases with increasing inlet temperature. Results have been verified with experimental data. These results indicate that a margin between several seconds and several hours (depending on the type of transient) is available before core damage may occur in an actual reactor.     

Small-scale component experiments of the penetration leak characterization test in the ALPHA program

A small-scale penetration leak characterization test has been performed as a part of the ALPHA program at Japan Atomic Energy Research Institute (JAERI). Two series of experiments were performed using test sections which simulate relevant parts of an EPA (Electrical Penetration Assembly) used in Japanese PWR containments. One of the test sections simulates an alumina module and the other includes the silicone resin portion of the EPA. The test section was heated in a leak test vessel which simulated thermal-hydraulic conditions inside and outside of the containment in a severe accident. From the experimental results, it was concluded that although the silicone resin may melt at high temperature, the alumina module will remain intact under severe accident conditions. The EPA as a whole is estimated to maintain leak-tightness during a severe accident. It was found in the experiments that heat conduction along the metal portion of the test section had a strong influence on the melt progression of the resin. It was also found that the measured strain of the alumina module was predominantly caused by the elevated temperature. Therefore, the thermal load will be more of a threat to the EPA's integrity rather than the pressure load.     

SCWR-FQT回路的冷却剂流量丧失事故研究

超临界水冷堆燃料验证实验（SCWR-FQT）将对1个小型燃料组件在超临界水环境下进行堆内性能测试。为了对该实验回路进行系统设计和安全分析,应用修改过的ATHLET程序建立实验回路计算模型,对两种造成燃料组件实验段冷却剂流量部分或全部丧失的设计基准事故进行模拟分析,即由于装载实验段的压力管内部的导向管破裂导致流经实验段的冷却剂旁通和主冷却剂泵卡轴事故。计算结果显示：实验段冷却剂旁通事故中,燃料包壳温度在事故初期出现约920 ℃的峰值;而主泵卡轴事故中,燃料包壳温度未明显升高。计算结果表明,现有的安全系统设计能保证在事故情况下维持燃料组件实验段的有效冷却。     

脉冲激光模拟SRAM器件单粒子翻转效应的试验方法研究

针对0.13 μm和0.35 μm工艺尺寸的两款商用SRAM器件进行了脉冲激光背部单粒子翻转效应试验方法研究。单粒子翻转效应主要测试单粒子翻转阈值和单粒子翻转截面,本文主要研究了激光聚焦深度、激光脉冲注量、测试模式和芯片配置的数据对测试两者的影响。试验结果表明：只有聚焦到芯片有源区才可测得最低的翻转阈值和最大的翻转截面,此时的结果与重离子结果基本一致;注量对翻转阈值测试无影响,但注量增大时翻转截面会减小,测试时激光脉冲注量应小于1×10⁷cm^-2;测试模式和存储数据对翻转阈值和翻转截面的影响不大,测试时可不考虑。     

矩形窄缝通道轴向非均匀加热临界热流密度试验本体设计

矩形窄缝通道轴向非均匀加热临界热流密度试验研试验数据处理等,而试验本体的设计是试验研究能否正常开展的关键.因此,准确、合理的试验本体的设计,对于矩形窄缝通道轴向非均匀加热临界热流密度试验研究是非常重要的.本文介绍了矩形窄缝通道轴向功率按截断余弦分布的临界热流密度试验本体的设计方法和结果.试验采用电加热方式,通过改变试验本体沿轴向的壁厚来实现非均匀加热,本文还介绍了试验本体的绝缘措施,临界测量方法等.     

Zircaloy cladding mechanical properties

This paper is concerned with life-limiting aspects of the Zircaloy cladding material, with emphasis on the fuel-clad interaction type of failure. The tensile test and creep properties and the changes of these due to neutron irradiation are reviewed. A section on high temperature properties of interest in loss-of-coolant accident (LOCA) analysis is also included. The paper is concluded with a discussion of the optimum choice of cladding material properties and the basis of the data reviewed.