ACP100集成式堆顶结构可压缩气体对流传热数值模拟

反应堆集成式堆顶结构的功能之一是通过冷却气体的对流传热为控制棒驱动机构提供冷却。针对ACP100反应堆集成式堆顶结构建立完整的模型并划分流场网格,基于ANSYS/CFX软件分别对可压缩气体和不可压缩气体进行计算,并严格对比分析了其结果,重点研究气体可压缩性对对流传热计算结果的影响。计算结果表明,气体可压缩性对速度、温度分布和压降有较大影响,忽略气体可压缩性所预测的控制棒驱动机构表面最高温度偏低,控制棒驱动机构间隙气体最大流速和压降也偏低。     

控制棒驱动机构自然循环冷却方式的可行性研究

传统的二代压水反应堆主要是采取鼓风机鼓风的方式对控制棒驱动机构(Control Rod Drive Mechanism,CRDM)进行强制通风冷却,该冷却方式耗能较大且无法保证绝对安全可靠。本文针对于大亚湾核电站中CRDM群的布置方式,采用中广核新型的EMC-B型控制棒驱动机构的结构及材料物性参数,运用了计算流体力学(Computational Fluid Dynamics,CFD)分析方法,研究了当采用空气自然循环冷却方式时,CRDM群及各线圈的温度分布,探索对CRDM群采用空气自然循环冷却方式的可行性。模拟分析结果表明:总体来看,处于外围和中心位置处的CRDM的线圈温度,要比中间区域的CRDM线圈温度高;对于给定计算工况,各线圈的最高温度为198°C,低于限制温度(200°C),表明对于所研究的CRDM群,依靠空气的自然对流,可以对CRDM进行有效冷却。计算结果可为新型CRDM群分布设计提供参考。     

基于华龙一号反应堆堆顶结构通风系统的数值模拟方法及试验验证

一体化堆顶结构对核反应堆中控制棒驱动机构(CRDM)的线圈组件通风冷却有重要作用。本文对华龙一号反应堆的一体化堆顶通风结构建立三维模型,运用ANSYS CFX模拟研究了在不同通风量下通风结构—CRDM线圈组件之间的气流速度和压降分布情况。为了验证数值模拟方法的正确性,建立1:1通风结构试验装置,在7×10~4、8×10~4、9×10~4和10×10~4m~3/h风量下进行冷态试验研究,试验结果与模拟结果吻合良好,验证了该模拟方法的正确性。考虑流体的对流传热,在控制棒驱动机构有热源的情况下模拟一体化堆顶结构的通风冷却性能。结果表明,一体化堆顶结构能很好地满足反应堆堆顶通风冷却的要求。     

控制棒驱动机构通风散热数值分析

利用ANSYS CFX软件对简化的CPR1000项目控制棒驱动机构三维模型进行连续下插动作时的通风散热数值模拟,分析控制棒驱动机构线圈温度、通风量与线圈温度、冷却风道进出口压差、控制棒驱动机构散热量及冷却风与磁轭和耐压壳接触面的传热系数的关系,并与试验结果进行对比,验证数值计算方法的可行性。     

快堆控制棒组件非均匀效应修正方法研究

控制棒组件是快堆控制系统和安全保护系统的重要组成部分,快堆控制棒价值的准确求解至关重要。基于PASC?5程序的快堆少群均匀化群常数计算中使用直接体积均匀化方式,这会导致控制棒价值严重高估,必须对控制棒组件的非均匀效应进行修正。基于群常数修正的思路,本论研究了体积?通量权重、反应率之比守恒和反应性守恒3种方法在快堆控制棒组件非均匀效应修正中的应用;基于二维特征线程序开发了群常数修正因子计算程序FRHP。通过中国实验快堆算例进行测试验证,修正后的控制棒价值计算结果与MCNP计算的参考结果符合较好,表明3种方法均能对控制棒组件的非均匀效应实现有效修正,其中反应性守恒方法修正效果最好。     

基于PLC的控制棒驱动机构复验

为了检验控制棒驱动机构的功能和性能是否满足设计规范书的要求,完成驱动机构的出厂试验,设计了控制棒驱动机构试验装置.基于可编程逻辑控制器(以下简称PLC)的控制棒驱动机构试验装置实现了控制棒的提升、保持、下降和报警综合等功能.文中介绍了控制棒驱动机构试验装置的原理和实现方法,通过和驱动机构的联调试验,证明控制棒驱动机构试...     

控制棒驱动机构步进运动特性研究

控制棒驱动机构步进运动由钩爪组件的提升衔铁在竖直方向上升和下降的交替运动实现。通过对控制棒驱动机构步进运动进行分解,采用有限元方法建立电路-磁路-机械运动耦合的动态计算模型,对控制棒驱动机构步进运动过程中电磁力、电流、位移和时间关系进行研究,得到步进运动的提升时间、下降时间、衔铁吸合时电流和电磁力等运动特性参数。     

金属骨架电磁线圈在控制棒驱动机构中的应用

对使用金属骨架电磁线圈的控制棒驱动机构(CRDM)与使用传统非金属骨架线圈进行性能比较。结果表明:虽然两者均符合CRDM设备规格书要求,但金属骨架在耐高温、耐辐射、抗震性能、机械强度、可加工性及经济性方面全面优于传统非金属材料骨架。采用金属骨架可以制作出耐温等级420℃及以上的耐高温CRDM电磁线圈,降低了对反应堆压力容器顶部CRDM通风冷却系统的要求。     

铀氢锆堆物理计算模型与程序

文中叙述基于两维四群中子扩散理论的铀氢锆堆物理计算模型及程序,以及用该模型计算的国外 TRIGA 堆的临界,控制棒效率等数据.     

上置式水力驱动控制棒

为提高200MW低温核供热堆经济性，对控制棒结构进行优化设计。在新的控制棒方案中，将控制棒驱动缸移到堆芯活化区以上，控制棒由浮动式活塞带动上下移动。由于驱动缸移出堆芯，燃料组件排布不再缺角，减小了堆的水铀比和堆内的中子吸收，增加了堆的运行时间。适当地加大驱动缸的直径和壁厚，有效降低了制造难度，提高了控制棒运行的可靠性。通过数值计算，分析了上置式水力驱动控制棒的落棒时间。     

示范快堆堆坑通风冷却三维数值模拟

为详细研究示范快堆堆坑内空气流动状态和温度分布情况,检验现行堆坑通风系统布置合理性与冷却效果,本文利用CFD软件对正常运行工况下的示范快堆堆坑空气流域进行三维数值模拟。结果表明,通风系统冷却效果满足设计要求,堆坑混凝土内壁最高温度为50.7 ℃,但堆坑内部流场复杂,温度分布的不均匀性较高,通风系统进出口排布方式需进一步优化。计算结果为主容器及贯穿件支承热工计算提供了更为准确的边界条件,为示范快堆一回路设计提供参考。     

控制棒驱动机构传热机理与隔热套性能研究

控制棒驱动机构(CRDM)依靠强制冷却措施维持工作温度。本文针对CRDM复杂的轴向传热机理,基于冷热侧流动的假设建立热虹吸自然对流分析模型,计算得到轴向温度分布及隔热套内径与热虹吸传热量之间的关系曲线;同时进行验证试验,测量不同情况下CRDM内外轴向温度分布和总散热量。通过分析和试验对比证明:基于假设的分析模型能模拟实际情况,热虹吸传质传热是CRDM轴向传热的主要途径,设置隔热套能有效抑制热虹吸、减少散热量。     

核反应堆控制棒驱动机构电机磁偏特性分析研究

核电厂反应堆的控制棒驱动机构（CRDM）采用同步电机作为机电能量转换的关键部件,电机的单边径向磁拉力会导致电机转子轴系变形,并加剧轴承的磨损,对CRDM寿命和核反应堆运行可靠性产生重要影响。本文分析了CRDM电机径向静偏心、径向动偏心、倾斜偏心的机理,并建立了径向磁拉力、轴向磁拉力计算的数学模型,分析获得了径向磁拉力、轴向磁拉力的变化规律以及与偏心量的关系,结果表明CRDM电机的径向磁拉力相对轴向磁拉力显著较大,并且与转子中心偏心量呈线性正比关系,与转子两端最多倾斜偏心量呈非线性正比关系,所得结论能够为CRDM电机优化改进和轴系结构设计提供指导和依据。      

喷射泵对自然循环过渡特性影响分析

分析了喷射泵在压水堆－回路自然循环过渡过程中的作用以及在不同流动条件下的阻力特性。分析结果表明：选择结构合理的喷射泵，可以改善压水堆一回路的过渡特性和自然循环能力；强迫循环条件下；压水堆一回路主循环泵有效压的损失随喷射泵阻力系数的增加而增加；自然循环条件下，喷射泵流动阻力系数影响压水堆一回路过度过程时间及自然循环流量的大小。为了改善压水堆一回路过度特性和提高一回路自然循环能力，可以采用无扩散段形式     

Wave propagation visualization in an experimental model for a control rod drive mechanism assembly

Nondestructive inspection techniques such as ultrasonic testing, eddy current testing, and visual testing are being developed to detect primary water stress corrosion cracks in control rod drive mechanism (CRDM) assemblies of nuclear power plants. A unit CRDM assembly consists of a reactor upper head including cladding, a penetration nozzle, and J-groove dissimilar metal welds with buttering. In this study, we fabricated a full-scale CRDM assembly mock-up. An ultrasonic propagation imaging (UPI) method using a scanning laser ultrasonic generator is proposed to visualize and simulate ultrasonic wave propagation around the thick and complex CRDM assembly. First, the proposed laser UPI system was validated for a simple aluminium plate by comparing the ultrasonic wave propagation movie (UWPM) obtained using the system with numerical simulation results reported in the literature. Lamb wave mode identification and damage detectability, depending on the ultrasonic frequency, were also included in the UWPM analysis. A CRDM assembly mock-up was fabricated in full-size and its vertical cross section was scanned using the laser UPI system to investigate the propagation characteristics of the longitudinal and Rayleigh waves in the complex structure. The ultrasonic source location and frequency were easily simulated by changing the sensor location and the band pass filtering zone, respectively. The ultrasonic propagation patterns before and after cracks in the weld and nozzle of the CRDM assembly were also analyzed. Since this visualization method is not limited in the flat cross section, it will be useful in developing ultrasound-based structural health monitoring technologies, advanced nondestructive methods, and numerical models. In addition, the proposed laser UPI system could be a useful tool in optimizing the receiver and transmitter locations, the ultrasonic path, and the ultrasonic frequency.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

单侧部分开口方腔自然对流数值模拟

为研究开口度对自然对流和传热的影响,采用有限容积法对单侧部分开口（2个开口）方腔自然对流进行了二维层流稳态数值模拟。计算结果表明：Ra和开口度的改变对方腔内的温度场分布和涡流结构有显著影响。相同开口度下,平均Nu和无量纲流量随Ra的增加而增加;相同Ra下,无量纲流量随开口度的增加而增加。当Ra>10⁴,可能存在最佳开口度,此时平均Nu取得最大值。     

中国一体化反应堆核电厂创新安全壳设计研究

中国一体化反应堆核电厂(CIP)是中国核反应堆系统设计技术国家重点实验室正在开发的新一代革新型、完全一体化的压水堆,其电功率约为300 MW.CIP采用堆内一体化布置,反应堆冷却剂系统设备以及控制棒驱动机构全部布置在反应堆压力容器内.这种一体化设计消除了传统的冷却剂回路管道,消除了大LOCA事故,具有更高的安全性.本文介绍了CIP安全壳系统方案选择、安全壳设计、安全壳设计压力的确定以及安全壳结构的计算分析.     

Verifications of the design calculations of the immersed cooler system during the start-up tests of SNR-300

The sodium cooled fast breeder reactor SNR-300 in Kalkar is equipped with two redundant immersed cooler systems (ICS). They are used only after the occurrence of serious accidents such as loss of all three main heat sinks. As part of the start-up tests of the reactor, the cooling capacity of the ICS, including natural convection behaviour, was verified and compared to the design data and calculations of the thermohydraulic computer codes NOTUNG and NANO. Three tests were performed. The first test, natural circulation in the sodium loops and air stack of the ICS, and the second test, forced convection in the sodium loops and natural convection in the air stack, were carried out to demonstrate the natural convection potential, even though this is not a design criterion. As part of the commissioning programme the third test with forced convection in the ICS was performed to prove the designed cooling capacity. During all three tests in-vessel cooling was based on natural convection.     

Some Relations of Neutronic Noise with Fluctuation of Inlet Coolant Temperature and Vibration of a Control Rod Obtained by Simultaneous Measurements at KUR

Experiments on reactor noise were conducted at KUR. Depending on the operating condition of the reactor, the cause of the noise are classified into the following four types.

1. Zero-power noise source due to the branching process of fission neutrons and/or due to random bombardment of neutrons to the detector—under natural circulation of coolant and at essentially zero-power level.

2. Coolant temperature fluctuation due to natural convection—under natural circulation and at relatively high power level.

3. Flow induced vibration of shim control rods—under forced circulation of coolant and at low power level.

4. Fluctuation of inlet coolant temperature—under forced circulation and near the maximum power level.

Vibration of a spare shim control rod and fluctuation of inlet coolant temperature were measured simultaneously with the neutronic noise. Then the noise sources of the types (3) and (4) were verified. The vibration of a control rod has a broad spectrum in low frequency region besides the large peak at 14 Hz. The fluctuation of inlet coolant temperature is non-white noise and consists of large low frequency component. The theoretically predicted sink structures in the neutronic PSD relating to the transit time of inlet coolant temperature fluctuation through the core were not observed in the experimental results.