针对突发断电事故下的核主泵惰转工况,基于额定参数提出惰转转速与惰转流量特性曲线计算模型,并通过100D型核主泵惰转试验数据对推导的模型予以验证。结果表明,该计算模型可用于核主泵初步设计计算和验证分析。基于该模型进一步得到了核主泵惰转设计准则,并验证了AP1000核主泵设计转动惯量。

For a sudden loss of power in reactor coolant pump (RCP) ,a calculation model of rotor speed and flow characteristics based on rated parameters was studied . The derived model was verified by comparing with the power-off experimental data of 100D RCP .T he results indicate that it can be used in preliminary design calculation and verification analysis .Then a design criterion of RCP was described based on the calcula-tion model .The moment of inertia in AP1000 RCP was verified by this criterion .

针对核电站主冷却剂泵故障特征微弱难以有效辨识及DST仅能解决在无高冲突辨识框架下的单一故障诊断等问题,提出一种基于DSmT决策级主冷却剂泵并发故障融合诊断模型.采用核主泵自由DSm模型和混合DSm模型对含有故障信息的多个独立证据源进行动态融合计算;分析核主泵DSmT故障特征信度赋值变化,确定主冷却剂泵故障(并发故障)诊断总决策.结果表明,将核仿真机采集TS、SS、VS和DS多源传感器数据直接对基本概率函数进行赋值,得出主冷却剂泵故障(并发故障)决策结果与实际工况相符,实例验证了所提方法的可行性、有效性及准确性.

Aiming at the problems that the faults of NPP ’ s main coolant pump are difficult to recognize effectively during their early phases, and that DST can only solve single fault diagnosis in the identification framework without serious conflict, this paper proposes a diagnosis model for main coolant pump’ s concurrent fault fusion in early phases, based on DSmT. Taking advantage of DSmT, multiple independent proofs are used to conduct fusion decision, and obtain the results of main pump fault and concurrent fault diagnosis. On the simulation machine of nuclear power plant, main coolant pump was simulated and verified, and the results show that this method can improve the accuracy of recognition of main pump ’ s current faults during early phases, and reduce the uncertainty of the diagnosis.

为研究核主泵模型泵导叶非均布对于其外特性及压力脉动影响，分别采用SST（剪切应力输运） k-ω湍流模型和分离涡模拟方法对泵内部流场进行了三维定常和非定常数值模拟，得到两模型泵外特性曲线和内部压力脉动情况，并对压力脉动进行时域频域分析。结果表明：采用特定形式的非均布导叶可以改善出口流动，提升模型泵多工况性能；导叶非均布对于泵内不同区域压力脉动影响不同，但其改变了由动静干涉产生的脉动频率分布，削弱了脉动幅值，有助于降低泵的振动和噪声，提高核主泵的安全性。

The shear stress transport ( SST) k-ωturbulent models and detached eddy simulations were used in steady and un-steady 3-dimensional numerical simulation to investigate the influence of uneven guide blade on performance and pressure fluctua-tion of nuclear model pump. Two model''s performance curves and pressure fluctuation were gotten,and both time and frequency domain of pressure fluctuation were analyzed. It is shown that,the flow condition at the outlet is improved with a specific configu-ration of uneven guide blade,and the performance of multiple operating conditions is promoted. Effects of uneven guide blade on pressure fluctuation in different area of model pump are of their own features. But the frequency and amplitude which Caused by rotor-stator interaction are changed and decreased. It helps to reduce the vibration and noise,improve the safety of nuclear main pump.

采用流场分析软件ANSYS CFX对核主泵在不同空化工况下的水动力特性进行数值模拟分析,利用Morlet小波变换和快速傅里叶变换对相应数据进行处理分析。结果表明:气体含量随压力的降低或时间的增加呈现出指数变化规律。在空化初生工况,核主泵扬程脉动频率以低频为主,叶轮流道内的压力脉动的主频仍以转频为主,而空化产生的压力脉动对主频的影响不明显。随着空化的发展,空化所诱发的压力脉动对主频、次主频及脉动幅值的影响越来越大,其扬程脉动频率以低频脉动为主。空化严重工况时,扬程脉动频率以无规律变化的脉动高频为主,同时包含近乎规律变化的脉动低频。

T he reactor coolant pump hydrodynamic characteristics at different cavitation conditions were studied by using flow field analysis software ANSYS CFX ,and the cor-responding data were processed and analyzed by using Morlet wavelet transform and fast Fourier transform .The results show that gas content presents the law of exponential function with the pressure reduction or time increase . In the cavitation primary condition ,the pulsation frequency of head for the reactor coolant pump is mainly low frequency ,and the main frequency of pressure pulsation is still rotation frequency w hile the effect of the pressure pulsation caused by cavitation on main frequency is not obvi-ous .With the development of cavitation ,the pressure fluctuation induced by cavitation becomes more serious especially for the main frequency ,secondary frequency and pulsa-ting amplitude w hile the head pulsation frequency is given priority to low frequency pulse .Under serious cavitation condition ,the head pulsa

为研究核主泵在变流量过渡过程中叶片数与导叶片数对径向力的影响，应用计算流体力学软件CFX对核主泵分别改变叶片数与导叶片数时，叶轮与导叶所承受的径向力变化规律进行数值模拟计算与试验，将数值模拟结果与试验进行验证，试验数据与数值模拟数据在误差控制范围之内。结果表明：变叶片数时，随着流量增大，核主泵叶轮所承受的径向力值随叶片数的增加而增大，在叶片数为7片时达到最大值，叶轮所承受的径向力的周期性趋于规律性；径向力的平衡性趋于最佳工况。随流量变小时，叶轮所承受的径向力位置随旋转周期有规律性地偏移和旋转，其偏移和旋转的变化梯度不同，叶轮所承受的径向力偏移和旋转的变化梯度明显大于向大流量增加时所对应的变化梯度。通过综合对比可知：在变流量过程中，当叶片数为5片、导叶片为11片时，叶轮承受的径向力是最小。

To study the effect of number of blades and guide vanes on a reactor coolant pump’s radial force under variable flow transition conditions,the computational fluid dynamics software CFX was applied in the numerical simulation and test for the radial force variation law with number of blades or guide vanes changed.Compared the numerical simulation results with the experimental ones,both of them were within the error control range.The results showed that with flow rate increasing,the radial force of the reactor coolant pump impeller increases with increase in number of blades,and reaches the maximum value with the blade number of 7,the periodicity of the radial force tends to regularity, the radial force balance tends to the optimum working condition;with flow rate decreasing,the position of the impeller radial force offsets and rotates regularly in its rotating cycle,and the variation gradients of offset and rotation are different;the variation gradients of the impeller radial