Real time reactor noise diagnostics for the Borssele (PWR) nuclear power plant

After the upgrade of Borssele NPP in 1997, core cycle 24, the power plant operated three years more with 91% availability. The authority of the power plant decided to enhance and upgrade the reactor trend monitoring and plant information recording system with higher frequencies than the plant data processing system (PPS) as well as installing a flexible and multiple-purpose reactor noise analysis system which may support the reactor maintenance group with on-line and off-line capabilities for several different signal processing applications. Two measuring and monitoring systems were built in 2001 and fully taken in implementation during the start-up of the new core 28. In this sense, the new system was used in power operation during the 29^th of September 2001. This paper will introduce the measuring system, the operational tasks, and the results obtained so far on the real-time core-barrel motions (CBM) and the two-primary coolant pump vibrations measured through the reactor noise analysis.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

Reactor noise measurements in the safety and regulating systems of CANDU stations

Reactor noise measurements of safety and regulating system intrumentation are performed in the CANDU nuclear power stations of Ontario Power Generation (OPG) and Bruce Power. Station signals included in the noise measurements are in-core flux detectors (ICFD), ion chambers (I/C), flow transmitters, pressure transmitters, and resistance temperature detectors (RTD). Their frequency dependent noise signatures are regularly measured during steady-state operation, and are used for parameter estimation and anomaly detection.

The specific applications include the following areas:

Flux noise measurements to detect and characterize (a) anomalies of in-core flux detectors, ion chambers and their electronics, (b) mechanical vibration of fuel channels and in-core detector tubes induced by coolant/moderator flow.

Pressure and flow noise measurements to estimate the in-situ response times of flow/pressure transmitters and their sensing lines installed in the reactor's coolant loops.

Temperature noise measurements to estimate the in-situ response times of thermal-well or strap-on type RTDs installed in the reactor's coolant and moderator loops.

Finite-element based vibration analysis of WWER-440 type reactors

A finite-element model describing the mechanical vibrations of the whole WWER-440 primary circuit was established to support the early detection of mechanical component faults. A special fluid–structure module was developed to consider the reaction forces of the fluid in the downcomer upon the moving core barrel and the reactor pressure vessel. This fluid–structure interaction (FSI) module is based on an approximated analytical 2D-solution of the coupled system of 3D fluid equations and the structural equations of motions. By means of the vibration model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated. It is shown that the FSI strongly influences those modes that lead to a relative displacement between reactor pressure vessel and core barrel. Moreover, by means of the model the shift of eigenfrequencies due to the degradation or to the failure of internal clamping and spring elements was investigated. Comparing the frequency spectra of the normal and the faulty structure, it could be proved that a recognition of such degradations and failures even inside the reactor pressure vessel is possible by pure excore vibration measurements.     

基于PCA的主泵传感器状态监测模型

核电厂主泵的主、辅系统中布置了大量的传感器,随着主泵的运行,传感器会出现不同程度的老化或故障。为了改善现有核电厂传感器周期性测试和校准方案的不足,提高运行的安全性与经济性,采用主成分分析（PCA）技术对主泵的传感器进行状态监测。使用某核电厂主泵的运行数据建立PCA监测模型,并利用该模型对传感器的小漂移故障和共模故障进行识别,仿真结果表明该模型对主泵传感器具有很好的监测效果。      

Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the subassemblies with high precision.

In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift.

The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow.

Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power.     

Conceptual design of a computer-driven display for monitoringreactor coolant mass

The conceptual design of a display for monitoring reactor coolant mass in the primary coolant loop of a pressurized water reactor is discussed. A means-end relational network serves as the framework for decomposing the abstract-the conservation law of mass-into system and process variables. A computer-driven display synthesizes, in real time, mass sources, mass inventory, and mass sinks based on measured plant variables. The purpose of the display is to aid human users in the cognitive tasks of monitoring operations, diagnosing system faults, and responding to component and system failures that affect mass     

An improved method for reactor coolant pump abnormality monitoring using power line signal analysis

An improved method to detect the reactor coolant pump (RCP) abnormality is suggested in this work. The monitoring parameters that are acquired from power line signal analysis are motor torque, motor speed and characteristic harmonic frequencies. The combination of Wigner–Ville Distribution (WVD) and feature area matrix comparison method is used for abnormality diagnosis. For validation of the proposed method, the test was performed during cool-down phase and heat-up phase in nuclear power plant (NPP) by cross-comparison with RCP vibration monitoring system (VMS). Using pump internal inspection results, the diagnosis prediction is verified.     

AP1000核电厂反应堆冷却剂系统布置设计

AP1000核电厂反应堆冷却剂系统布置设计,在满足系统功能的前提下,充分考虑了屏蔽防护、核级部件在役检查、模块化设计、内部灾害防护等方面的要求。反应堆冷却剂系统主设备及主回路采用了紧凑型的布置方式,改善了环路配置的经济性,波动管布置在考虑足够柔性的基础上采用了大倾斜角连续上坡的方式,降低了波动管在运行过程中出现热分层的可能性,稳压器安全阀及ADS第1、2、3级集中布置在稳压器顶部,组合成一体化的模块Q601,改善了反应堆冷却剂系统布置结构。     

压水堆核电站一回路工况变化对主泵主要机械性能的影响

论述了大亚湾和岭澳1000MW压水堆核电站反应堆冷却剂回路(一回路)主要瞬态工况对反应堆冷却剂泵的主要机械性能参数的影响,为避免主泵受瞬态干扰,以及通过改变系统参数调整来改善主泵机械参数提供了理论依据。     

三门核电站停机不停堆的运行分析

三门核电AP1000反应堆在满功率情况下发生汽轮机故障停机事件时,通过快速降功率系统、旁排系统和棒控系统等的快速响应,一回路的参数不会突破安全限值,避免了反应堆停堆,降低了该瞬态对反应堆冷却剂系统的冲击。文章对停机不停堆的实现方式和运行特点进行了详细的分析和阐述,以帮助电站人员对停机不停堆的理解,并提高他们面临瞬态的响应能力。     

Heats of Vaporization of UO2 and UO from UO2

In an earlier paper, a stochastic model of a power reactor has been proposed by the present author on the premise that the coolant-flow through a core is usually accompanied by random variations in the flow-rate, which are eventually largely responsible for the internal reactivity fluctuations.

In the present work, this model is extended to three different reactor systems: (a) where there exists a relaxation process corresponding to the effect of buoyant flow; (b) where a control or fuel element vibrates randomly, due to coolant flow-rate fluctuations; (c) where there are fluctuations in the inlet temperature with a non-white spectrum.

The noise spectra are derived for various state quantities with use made of the Langevin procedure. The theory is illustrated by referring chiefly to the neutron noise spectra, and comparing with the results of observations.

It is shown that the noise sources in question contribute significantly to the spectra, as compared with a low frequency component due to an inherent noise source in the coolant flow. In particular, a strong resonance peak of the spectra arises from the coupling between the random mechanical vibrations and coolant the flow-rate fluctuations.     

Development of an acoustic leak monitoring system

The purpose, structure, and basic characteristics of an acoustic system for monitoring coolant leaks in the first loop of a VVER reactor are presented. The principle of operation of the diagnostics algorithm is described. The system has been checked on a special stand. The system satisfies all sensitivity and temporal requirements. In 2005, SAKT was put into experimental operation in the No. 3 unit of the Kalinin nuclear power plant. __________ Translated from Atomnaya énergiya, Vol. 103, No. 6, pp. 342–347, December, 2007.     

Measurement of subassembly outlet coolant temperature in the JOYO experimental fast reactor

Thermocouple temperature sensors are installed above the central region of the core in the JOYO experimental fast reactor to monitor the outlet coolant temperature of 115 subassemblies. This paper summarizes the experimental temperature data obtained during initial 50 MWt operation of the reactor. Subassembly outlet coolant temperature distributions that were obtained under various power levels, different main cooling system flowrates, and unequal reactor inlet temperatures from the two cooling loops are described. In addition, coolant temperature and flowrate distributions at the subassembly outlet measured in a zero power experiment are presented.     

反应堆冷却剂泵转速测量系统改进设计及实践

反应堆冷却剂泵（主泵）转速是核电站关键设备反应堆冷却剂泵运行状态监测的重要参数，直接反映设备的运行状况，并担负向反应堆保护系统输送反应堆的保护信号。但是该信号一直存在运行过程中测量不稳定的情况。从该测量通道的测量原理、历史状态，结合现场的实际检修过程，对转速测量的缺陷、可能的原因进行分析，同时对以上原因采取改进方式。经过2008年的运行验证，改进的测量方式信号稳定，满足了现场的要求，有利于改进的持续进行。     

反应堆冷却剂泵动压机械密封的工程开发与应用

反应堆冷却剂泵（以下简称主泵）轴密封由3级相同的动压机械密封串联组成，是主泵的心脏，其泄漏量直接决定主泵能否正常运行。本文提出了一种新型的挤压变形研磨法完成动压机械密封的制造，应用挤压变形工装和金属垫片使静环产生变形，在密封端面研磨出9个波形槽。功能实验表明，新型的机械密封在考核压力下的低压泄漏量满足主泵轴密封的设计要求；压力突变工况下的冲击考核实验表明，新型的动压机械密封摩擦副之间的液膜刚度未发生破坏，未出现密封失效。本文研发的动压机械密封在核电厂的运行状况与实验结果完全吻合，充分证明了该新型动压机械密封具有极高的工程应用可靠性。      

Borssele PWR noise: Measurements, analysis and interpretation

In the Borssele reactor — a 450 MWe PWR — reactor noise measurements have been performed during four fuel cycles. Measurements were made with a set of ex-core neutron detectors, on one occasion an in-core displacement transducer, and with primary coolant pressure sensors. Digital analysis was applied, where the most powerful tool was the computer programme FAST, which computes auto and cross power spectra for all combinations from a set of many simultaneously recorded signals.

Analyses of neutronic signals show a reactivity noise peak at 9.2 Hz, core barrel motion peaks at about 12 and 15 Hz, a damped oscillation at about 2 Hz. Results are given for begin and end of each fuel cycle. The r.m.s. value of the low frequency noise appears to depend linearly on the boron concentration over a wide range.

Also some results of primary coolant pressure noise are presented, with coherent peaks below 15 Hz and incoherent peaks above.

The second part of the paper describes an alternative way of analyzing and interpreting noise spectra, namely attempts to decompose the neutronic power spectra into physical components, using the information present in the CPSD's of all detector combinations. The components are characterized by their detector position dependency. Effects considered are: uncorrelated noise, global reactivity noise, core motion attenuation noise, and a possible coupling between reactivity and core motion. Results show excellent separation into reactivity and core motion components with possibilities to separate overlapping peaks. Weak peaks become more easily detectable. At low frequencies the decomposition of the spectra is not yet complete, however.     

Performance diagnostics of the thermoelectric circuits of a system of in-reactor monitoring

Conclusions The results obtained by TC diagnostics, monitoring the coolant temperature at the pile outlet over the course of reactor operation at power, indicate the benefits of using diagnostics — a simple, reliable, and at the same time objective instrument for detecting the presence and character of faults. This is very important not only for the technical maintenance of the TC but also in increasing the reliability of the measured temperature data required for monitoring the reactor operation. In reactor operation using diagnostics, TC faults may be automatically traced, and TC faults which remain unremarked with existing monitoring methods may be detected.In conclusion, thanks are offered to all our colleagues and, most of all, to S. Badyar, L. Arbet, and D. Zhizhka; and also to the service personnel of the atomic power plant for their active participation in the work of developing and introducing the diagnostics.Research Institute of Atomic Power Stations, Yaslovskii Bogunitse, Czechoslovakia. Translated from Atomnaya Énergiya, Vol. 52, No. 4, pp. 244–248, April, 1982.