5×5花瓣形燃料棒束组件内单相流动与换热特性数值模拟研究

相较于传统圆柱形燃料棒,花瓣形燃料棒具有安全裕量高等优点,研究其在压水堆运行工况下的热工水力特性具有重要意义。本文通过STAR-CCM+对5×5花瓣形燃料棒束组件进行数值模拟研究,计算并分析了组件内二次流速度、温度、换热系数等关键热工参数,获得了入口流速、螺旋节距对组件内部流动与换热特性的影响规律。计算结果表明：花瓣形燃料棒的螺旋结构可增强冷却剂的横向流动,同一高度上燃料棒表面温度分布具有周期性,增大入口流速可增强燃料棒的表面换热,消除温度分布的不均匀性。此外,螺旋节距大于750 mm,燃料棒换热性能与无扭转的燃料棒相差不大,甚至更低。     

先进工艺UO2芯块堆内性能试验研究——燃料组件水力试验

为了确保试验燃料组件的辐照安全，需确定流经燃料棒冷却剂的流速，验证燃料辐照装置的设计能否满足试验要求。实际的辐照装置因条件限制不带流量测量仪表，所以，对燃料组件进行堆外、堆内水力试验，并根据测量结果对流经辐照装置的冷却剂流量进行推算。     

带7道格架的5×5棒束两相性能CFD分析

采用两相计算流体动力学（CFD）方法进行带7道格架的5×5棒束两相性能研究，其中结构搅混格架（MG）和跨间搅混格架（MSMG）交替布置，计算考虑汽泡合并与破裂、热量传递，但不考虑相间的质量传递。为选择合理的两相模型参数，首先以带2道格架（MG、MSMG）的AFA3G燃料组件5×5棒束架为研究对象，对最大气泡直径、汽泡合并破裂系数、非曳力模型及曳力模型、入口气泡直径、入口空泡份额分布等进行了敏感性及不确定性分析。此后采用该两相模型设置，针对带7道格架的AFA3G燃料组件进行了两相性能研究，计算结果显示格架间的各项参数不存在完全一致的周期性，但同种格架上游的空泡份额分布具有一定的相似性，因此用于两相性能评价可计算带2~3道格架的棒束，该研究可用于带格架棒束两相计算的模型设置与几何规模选择，为下一步采用两相CFD计算建立燃料组件热工水力性能评价准则奠定了基础。最后比较了AFA3G燃料组件及改进型燃料组件两种格架的空泡分布特性，并从提高燃料组件临界热流密度（CHF）特性的角度对其进行评价，获得与实验一致的结论，证明了评价方法的正确性。      

燃料组件结构参数对于控制棒组件落棒缓冲效果的敏感性分析

压水堆控制棒落棒水力缓冲过程主要依靠导向管下部缓冲段结构实现。燃料组件缓冲段结构参数的选取对于落棒缓冲效果影响较大。基于水力学基本公式构建的落棒缓冲理论模型,运用数值计算的方法研究缓冲段结构参数,如导向管与控制棒环形间隙、轴肩螺钉流水孔径与长度等对落棒水力缓冲效果的影响情况。总结出落棒缓冲影响规律,可作为燃料组件导向管结构设计的参考。     

基于克里金方法的快堆燃料组件设计

为详细研究快堆组件棒束中的流动与换热两方面因素对组件热工水力特性的影响，本工作采用克里金方法研究快堆燃料组件的设计参数。由计算结果可知：保证组件出口平均温度不变，随组件压降的升高，满足条件的P/D和H/D范围变化有一定的方向性，逐渐靠近原点；保证组件棒束的压降不变，随组件出口平均温度的升高，P/D和H/D范围变化不具备方向性。根据计算结果可在给定输入限值条件下得到组件满足条件的设计参数范围，可为今后大型快堆的燃料组件选型提供参考。     

基于CFD方法的行波堆燃料组件燃烧区热工流体特性研究

采用计算流体力学（CFD）方法对行波堆燃料组件7棒束、19棒束及37棒束模型进行计算分析,发现行波堆燃料组件内冷却剂温度随轴向高度增加逐渐升高的同时具有逐渐向中心区域聚集的效应,组件出口区域垂直于流动方向的截面冷却剂温度分布差别很大,对边距约为26 cm的组件中心区域与外围区域最大温差超过100 ℃。组件内较大的冷却剂温度梯度主要出现在组件最外两圈燃料棒及组件盒之间的区域,而其他区域温度梯度较小,该结论可初步推广到有217根燃料棒的行波堆燃料组件。现有行波堆燃料组件结构需进一步优化。     

压水堆燃料组件横向非线性特征模拟研究

压水堆燃料组件的燃料棒依靠格架进行夹持,保持燃料棒的横向和轴向定位。在燃料组件弯曲时,燃料棒与格架产生相对滑移,是燃料组件产生横向非线性特征的主要原因。本文分析典型的压水堆燃料组件格架和燃料棒夹持系统的设计特点,结合分析和试验结果,将夹持系统的滑移和弯曲特性分解为滑移单元和旋转弹簧单元的效应,从而实现夹持系统的力学模拟。通过计算与试验结果比较,验证所建立的夹持系统模型的有效性。将夹持系统模型用于燃料组件横向非线性模型中,通过模型计算与燃料组件横向拉伸试验结果对比,符合性良好。     

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。     

300MW燃料组件定位格架导向翼三维流场分析

从三维流场分析角度验证了300MW燃料组件定位格架导向翼初步设计改进方案的合理性,采用流体计算软件对燃料组件定位格架或棒束定位格架进行计算,比较导向翼修改前后的定位格架水力特性和导向翼周围流场的变化,为实际工程应用提供了依据。     

超临界水冷堆燃料棒性能分析程序适用性研究

根据超临界水冷堆(SCWR)燃料棒的热工水力特点,基于压水堆(PWR)燃料棒性能分析程序的理论模型和计算方法研究燃料包壳的物性模型和超临界水(SCW)与燃料包壳的传热模型,建立适用于SCWR燃料棒的性能分析程序——SCWRFPA。采用SCWRFPA和可分析SCWR的热工水力子通道程序ATHAS分别对1/8欧洲超临界轻水堆(HPLWR)燃料组件燃料棒进行计算,其计算结果基本一致。     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

加/减速流动下棒束通道内速度分布和湍流特性研究

事故工况及海洋条件下反应堆处于非稳态工况,堆芯燃料组件内热工水力行为复杂多变,对反应堆安全提出了更高挑战,因此有必要对非稳态下燃料组件内流动换热特性开展研究。基于粒子图像测速（PIV）技术,结合远心镜头和脉冲控制器,实现对燃料组件内复杂流场的高时空分辨率、长时间的连续测量,获得了流量波动下燃料组件内时空演变的流场结构,分析了棒束通道内速度分布、湍流强度、雷诺应力等瞬时流场信息的空间演变特性。以定常流动下流场分布特性为基准,对比分析了加速度对燃料组件内空间流场分布的贡献特点。实验结果表明：加速流动提高了棒束通道内流层之间的速度梯度,抑制了横向速度和湍流强度;减速流动减弱了棒束通道内流层之间的速度梯度,提高了横向速度和湍流强度。实验结果有助于揭示燃料组件在非稳态条件下的瞬态特性,并为燃料组件的设计和优化奠定基础。     

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.     

中国先进研究堆标准燃料组件堆外水力稳定性试验

中国先进研究堆（CARR）标准燃料组件由滚压在两块侧板上的21块燃料板组成。堆外水力试验的目的是考验在水力冲刷条件下燃料组件的结构稳定性。试验件是按照正式产品制造工艺制造的贫铀组件，试验平均流速为12m/s，是满功率运行流速的120％。先后试验了2个组件，第1个组件试验60d，是满功率运行时间的120％，试验后观察到固定下定位梳的销钉松动，下定位梳严重磨损了燃料板;工艺改进后制造的第2个组件试验120d，是满功率运行时间的240％，试验表明，第2个组件结构完整。试验中对组件结构稳定性和燃料板腐蚀性能，诸如组件的压差、燃料板振动、包壳表面腐蚀深度等进行了研究。     

Statistical evaluation of pin gaps in fuel clusters for fugen (II)

Fuel pin gaps of Fugen fuel assemblies deviate statistically from their nominal value due to manufacturing and assembling tolerances which influence the thermal and hydraulic characteristics of the reactor core. For assurance of the minimum fuel pin gap, an analytical method of evaluating the reliability of spacer gauge tests applied to selected fuel pin gaps arrayed within a Fugen fuel assembly is discussed where a computer program STGAP is utilized.Correlations among the thickness of a spacer gauge, the reliability of the test and the rate of rejecting fuel assemblies whose pin gaps all satisfy the minimum design criterion are discussed in connection with the optimum gauge thickness for a given realiability level of the test. Sample calculation shows that fuel subassemblies installed in a Fugen reactor core have the overall reliability level of 99.9954% at the beginning of fuel life.     

Experimental investigations of the hydraulic field in wire-wrapped LMFBR core assemblies

An investigation of the hydraulic behavior of wire-wrapped fuel and blanket assemblies was conducted in an air flow test facility. The test section was a large scale sector (slightly more than one-sixth) of prototypic fuel and blanket assemblies of the Clinch River Breeder Reactor Plant; the scale factor was approximately 11:1 and 5:1 for the fuel and blanket, respectively, thus allowing a very large number of measurements within each subchannel.The purpose of these experiments is discussed along with a brief state of the art review; also discussed is the role of these tests on the core thermal-hydraulic design through calibration and verification of the analytical codes employed in the design. The test section and experimental procedures are illustrated. Experimental results are discussed in detail: static pressure gradients; local and average cross flow through the gap spacing between rods as a function of the wire wrap position and at all typical locations in the assembly; detailed axial velocity mappings in the inboard and peripheral channels. The physical significance of the results is interpreted and the fundamental difference in the hydraulic behavior of fuel and blanket assemblies is pointed out, discussed and explained in terms of fundamental geometric parameters. The application of the fuel assembly data to calibration/verification of subchannel analysis and distributed parameter codes is presented in detail. A quantitative model of the cross flow driving forces is elaborated as the starting point for a comprehensive phenomenological modeling of the hydraulic behavior of wire-wrapped assemblies.     

Influence of fuel bundle loading errors on the subcritically during refueling campaigns for the present BWR cores of KRB-II

On the basis of real fuel assembly inventories as they are presently available in KRB-II, the influence of fuel bundle loading errors on the subcriticality during refueling campaigns was investigated with the calculational methods of the incore fuel management. To this, control rod cells which show the least shut-down reactivity were considered and less reactive fuel assemblies were successively exchanged with fuel assemblies of highest possible reactivity from distant core regions. The results show that the total shut-down reactivity is only reduced by a comparatively small amount. The stuck rod shut-down reactivity, on the other hand, is strongly diminished with increasing number of locally concentrated mislocated fuel assemblies of highest possible reactivity. Thus, unintentional criticality cannot be reached during refueling campaigns with all control rods inserted. In conjunction with the deliberate withdrawal of one control rod, two or three mislocated fuel assemblies can cause criticality, depending on the absolute value of the realized stuck rod shut-down reactivity.     

超临界水冷堆双排燃料组件子通道分析

研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.