控制棒驱动机构轴向传热特性试验和理论分析

核反应堆控制棒驱动机构（CRDM）是反应堆关键的安全设备,担负着反应堆的启动、功率调节和安全停堆等重要功能。为保证CRDM在运行工况下的温度,本文对某一型号CRDM轴向传热特性进行试验和理论分析。通过试验,测量磁轭线圈位置处驱动杆内壁温并建立试验系统加热功率与冷却风速的关系,获得了保证磁轭线圈正常工作的最小风速以及最小散热量。通过理论分析,建立了CRDM行程壳体内部热虹吸现象的传热模型,模型计算结果与试验结果吻合良好。     

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。     

高性能粘结Nd-Fe-B磁钢研究

为满足专用机器的设计需求,需研制大体积、高性能的各向同性磁钢.按照传统的烧结成型工艺生产的Nd-Fe-B磁钢,磁均匀性差,合格率低,不适宜大规模生产,因而成本高.本工作采用注射成型生产工艺,经不断研究改进,研制出新型XNA8磁钢.对其实行恰当的充磁,磁钢的轴向磁拉力、径向刚度及磁偏心能够全面满足设计要求.产品经批量考核后,确证其尺寸精度高,磁性能一致性好,易于批量生产,经济性好.     

控制棒驱动机构磁路饱和特性分析

利用理论分析和有限元分析（FEA）相结合的方法，研究了控制棒驱动机构（CRDM）中磁路饱和对电磁参数以及磁路参数的影响。通过建立CRDM等效磁路（EMC）模型，推导获取磁路参数的方法，并基于FEA获得了CRDM电磁参数以及磁路参数随电流和主气隙长度的变化趋势。分析结果表明，主磁路剩余磁阻和线圈电感受磁路饱和的影响较大；CRDM漏磁系数与磁路饱和的关系较小，但受主气隙长度的影响较大；主气隙磁阻的计算应考虑气隙磁通边缘效应。本研究为CRDM的EMC模型优化提供了理论依据。     

同步磁阻电机型控制棒驱动机构矢量控制技术研究

针对同步磁阻电机型控制棒驱动机构（CRDM）电磁高度耦合难以对输出力矩进行有效线性调节的问题,研究了同步磁阻电机型CRDM的矢量控制技术。将同步磁阻电机CRDM的数学模型映射至同步旋转坐标系内,对输出电磁力矩与磁链进行解耦。通过励磁电流与力矩电流的独立调节,实现输出力矩的线性控制。利用MATLAB/SIMULINK搭建了同步磁阻电机型CRDM的矢量控制模型,对控制方案进行验证。仿真结果表明,本研究的控制方案具有更优的响应速度、稳态精度以及平稳性。      

一种磁力载荷模拟器的设计和研究

在研制10MW高温气冷堆氦气透平发电系统(HTR-10GT)过程中,需开展透平压气机转子的动力学性能试验。为了模拟透平压气机转子在运转过程中受到的气动力和不平衡力,需对转子施加一定载荷。本文介绍了一种磁力载荷模拟器的设计,该载荷模拟器采用与磁力轴承相似的技术,通过电流产生磁力直接在模拟转子上加载轴向和径向载荷。磁力载荷模拟器包括轴向载荷模拟器和径向载荷模拟器两部分,轴向气动力载荷用轴向载荷模拟器模拟,叶片气动周向力载荷和不平衡力离心载荷用径向载荷模拟器来模拟。轴向载荷模拟器可施加载荷设计值为10kN,径向载荷模拟器载荷设计值为715N,气隙均为1mm。通过分析计算,磁力载荷模拟器完全可满足本试验中的加载要求。同时,磁力载荷模拟器可灵活加载,在线加载,快速改变载荷大小和形式,为电磁轴承的应用开发了一个新领域。     

氦气透平轴向磁力轴承承载力-转速的内在关系

本文研究了轴向磁力轴承承载力与转速、材料强度之间的内在关系。轴向磁力轴承采用控制电流所产生的电磁力,使转子维持在悬浮工作位置不变,因此其工作原理决定了它的承载力与推力盘面积成正比。通过对推力盘的强度分析,找出了最大切应力和径向应力的位置。分析发现,推力盘的面积受轮缘处最大切应力的限制,与材料许用应力成正比、与转速成反比。本文给出的推力盘最大承载力与材料强度和最高转速的内在关系可用于指导氦气透平压气机轴向磁力轴承的设计。     

驱动杆位置对控制棒驱动机构轴向传热特性的影响

核电站常采用对控制棒驱动机构(CRDM)外部强制通风冷却的方法来确保其正常运行。CRDM外部的通风冷却会引起内部的热虹吸现象,从而导致CRDM轴向温度升高。研究表明,驱动杆插入CRDM中的不同位置是影响热虹吸现象的重要因素之一。本文通过试验研究了多种风速下驱动杆在全高度棒位和半高度棒位时CRDM的轴向温度分布。结果表明,驱动杆在全高度棒位时CRDM内部的热虹吸现象更剧烈,冷却空气带走的热量更多。     

八万千伏安交流脉冲发电机组

本文综述了八万千伏安交流脉冲发电机组概况,在研制这类发电机中,对电磁方案设计、转子表面谐波发热温升判定、定子绕组和转子阻尼绕组的特殊结构、六相电机突然外部短路分析实验、轴系结构与扭振等六个特殊问题的处理技巧和方法进行了分析。结合主要的实验数据和几年运行情况,对八万千伏安交流脉冲发电机组进行了综合性评价。     

八万千伏安交流脉冲发电机组

本文综述了八万千伏安交流脉冲发电机组概况,在研制这类发电机中,对电磁方案设计、转子表面谐波发热温升判定、定子绕组和转子阻尼绕组的特殊结构、六相电机突然外部短路分析实验、轴系结构与扭振等六个特殊问题的处理技巧和方法进行了分析。结合主要的实验数据和几年运行情况,对八万千伏安交流脉冲发电机组进了综合性评价。     

Application of an integrated power line signal analysis methodology to the detection of major rotating machine abnormalities in nuclear power plants

An integrated and improved method to detect and identify the abnormality of motor driven rotating machinery in nuclear power plants (NPPs) using power line signal analysis is suggested in this work. The primary goal of this work is to improve the motor current signature analysis (MCSA) method that has been used as an alternative or supplement of the conventional vibration monitoring system (VMS). Through this work, the integrated system using both modulated flux density model (MFDM) and rotating flux model (RFM) is proposed. The MFDM is based on the fact that the major mechanical vibration of rotating machines can be normalized to the motor air-gap eccentricity and the modulation of air-gap flux density. Therefore, if the major defect such as bearing defect or the shaft deformation is present, it is identifiable through the power line signal resulting from the modulated magnetic density. Moreover, the broken rotor bar state or rotor eccentricity due to electrical imbalance can be analyzed using the RFM. The other important feature of this system is an automated abnormality detection and diagnosis algorithm. It is possible to diagnose the abnormality without relying on experts in NPPs. The verification is done through varying load/torque test experiment as well as via computer simulation in this work. The experimental results show that they are in good agreement with the simulated results.     

In-vessel Type Control Rod Drive Mechanism Using Magnetic Force Latching for a Very Small Reactor

A highly reliable control rod drive mechanism driven by an electric motor installed inside the reactor vessel (INV- CRDM) for a very small reactor has been designed. The INV-CRDM contributes to the compactness and simplicity of the reactor system, and can eliminate the possibility of a rod ejection accident. In the design, a new type of latch mechanism using an electromagnetic force to directly connect both of the shafts, one of which was the motor driven shaft and the other the control rod driving shaft, was applied so as to make the INV-CRDM very compact. The cable supplying current remained stationary, even when both of the shafts was moving. The required functions of the latch mechanism are to maintain an adequate latching force for the control rod shaft to move within a stroke of 370 mm, and to release the shafts in a shorter time than 0.2 s after a scram signal is received. A functional test with a model that approximately simulated the design was conducted to test the latching force and de-latching at room temperature. The test showed that the latching force increased with the current of the magnet coil, as did the de-latch time. The post-test analysis with a finite element analysis code revealed that the clearance between the two shafts greatly affected the latching force. With the same analysis method, the design analysis of the latch mechanism at a high temperature condition of 300°C was conducted, and it was confirmed that the latch mechanism contained enough latching force.     

Evaluation of alternate power conversion unit designs for the GT-MHR

General Atomics in the USA and Experimental Design Bureau of Machine Building (OKBM) in the Russian Federation have jointly developed a nuclear power plant design, the gas turbine modular helium reactor (GT-MHR). There have been considerable improvements during the last 10 years, which resulted in a more effective, efficient and safe design. The existing design is based on a 600 MW(t) reactor cooled by helium at a pressure of about 7 MPa. The power conversion unit (PCU) uses reactor outlet helium with a temperature of 850 °C in a direct Brayton cycle to achieve the cycle efficiency of about 48%. The PCU consists of a gas turbine, a recuperator, a precooler, low-pressure and high-pressure compressors, an intercooler, and a generator. The turbomachine (TM) includes the turbine, compressors and generator mounted on a single vertical shaft. TM shaft rotation speed is 4400 rpm. The shaft of generator is connected to the turbine shaft by a flexible coupling. The required grid frequency of generated electricity is achieved by a converter. All PCU components are enclosed in a single vessel. TM uses radial and axial electromagnetic bearings (EMB) for support. Catcher bearings (CB) are provided as redundant support for the TM rotor in case of EMBs failure. Several alternative PCU designs were analyzed on the basis of current progress in technologies, new world experience, and experience accumulated in the process of GT-MHR design development. Results of these analyses will be taken into account when a final PCU design is selected.     

控制棒驱动机构电流监测与故障诊断技术研究

控制棒驱动机构作为反应堆控制和核安全保护系统的执行机构，对其运行状态进行有效监测是防止反应堆发生控制棒卡棒、滑棒和驱动失灵的重要措施。本文在分析控制棒驱动机构动作原理的基础上，通过对驱动机构探测方法、信号分析处理方法以及故障鉴别方法的研究，实现了对控制棒驱动机构的电流监测与故障诊断，为驱动机构电流监测与故障诊断系统的应用奠定了技术基础。      

基于自然对流的控制棒驱动机构温场研究

在新一代反应堆控制棒驱动机构耐温等级提升后,为分析取消控制棒驱动机构强制通风系统是否可行,建立了基于单台控制棒驱动机构流场的计算模型,合理确定了求解方法和边界条件,并通过与试验结果进行拟合,进行了控制棒驱动机构群全尺寸模型的温场和速度场计算,得到基于自然对流的控制棒驱动机构温场效应。研究结果表明,通过自然对流可以将控制棒驱动机构运行产生的热量完全带出堆坑,不存在堆坑窝热情况,并且随着控制棒驱动机构群总输出功率的增加,堆顶区域的自然循环强度也增加,可以达到冷却控制棒驱动机构的效果。因此,耐温等级提升后的控制棒驱动机构可以取消强制通风。     

Model validation and higher order sliding mode controller design for a research reactor

The paper pertains to model validation and novel higher order sliding mode controller design for a nuclear research reactor. Sliding mode controllers for nuclear reactors were reported before but higher order sliding mode controllers have added advantage of reduced chattering. As a first step of model development a simulation model of control rod drive mechanism (CRDM) has been developed using SIMULINK^®. This model has been validated with a lab based CRDM model, which is similar to Pakistan Research Reactor-1 (PARR-1) CRDM system. The nonlinear model of PARR-1 has been tuned and validated with experimental data. This model has been subsequently used for higher order sliding mode controller design and performance evaluation. Certain parameter values have also been recalculated to ensure model accuracy. Based on the validated model a robust nonlinear controller for controlling output power by manipulating control rod position has been developed and simulated. The new controller showed improved performance as compared to the classical PID controller.     

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.     

偏心率对核主泵叶轮口环密封激励力的影响研究

为探究转子偏心率对核主泵转子密封激励力的影响,基于雷诺时均N-S方程和RNG k-ε湍流模型,选取平面密封、迷宫密封和螺旋密封3种口环结构方案,对核主泵口环间隙内部流动进行数值计算,得到口环间隙区域压力、泄漏量及其密封激励力的分布规律。结果表明,模型泵性能预测值和试验值较为吻合,扬程最大误差为4.78%。在转子无偏心时,相对于平面密封,采用螺旋密封方案时口环泄漏量显著降低93.1%,而密封激励力增加63%。偏心率为10%时,口环压力分布沿周向较为均匀;当偏心率为30%时,周向靠近偏心位置处,口环间隙内部产生带状压力突升区,相对于无偏心方案,平面密封的泄漏量显著降低43.6%,而密封激励力增大4.4倍,迷宫密封和螺旋密封方案可显著降低转子偏心产生的密封激励力,其中迷宫密封可显著降低55%;偏心率为50%时,口环间隙内部带状压力突升区域偏向高压侧。本数值预测方法为揭示偏心转子对核主泵口环密封激励力的影响提供理论依据。