基于单孔包层与薄膜组成FP干涉仪腔的新型光纤应力传感器研究

本文研究了一种新型基于单孔包层与薄膜组成Fabry-Perot(FP)干涉仪腔的光纤应力传感器,单孔的一个表面为45°以使光线传输从横向转到纵向,从而使一半的包层与有机薄膜组成FP干涉仪腔。采用频率多路复用的方法获得了11个传感器复用在一个系统的频谱。在采用合适的不同有机薄膜厚度的情况下,系统能集成更多的光纤传感器。研究结果表明该新型的基于单孔的光纤传感器在核工业领域实时监测方面具有潜在应用。     

基于光纤珐珀结构的耐高温应变传感器研究

为了研制一种适用于反应堆一回路高温环境下机械测量的微型应变传感器,基于光纤珐珀（FP）结构光学原理、解调原理和封装技术研究,提出传感器的热机耦合模型,设计出一种基于光纤FP结构的耐高温应变传感器及其解调系统,并对其在常温空气中和高温中压水环境中进行动态测试。研究和测试结果表明,该高温光纤FP应变传感器及其解调系统能够在高温、中压以及水环境下稳定工作,且应变传感器的测量工作范围为0～4000με,解调系统速度达到5 kHz。本研究设计的高温光纤FP应变传感器及其解调系统能够用于反应堆一回路高温环境下的应变测量,从而为反应堆一回路高温环境下的监测提供了一种有效的监测方法。     

光纤激光荧光法测定硝酸体系中微量铀

一、引言 微量铀的测量方法已有大量报道,但绝大部分的方法,一般需经化学分离,因此不适用于远距离的在线分析。近年来,随着光纤传感器的出现及激光荧光技术的发展,远距离在线分析Purex流程中铀和钚的方法引起了人们的兴趣。J.R mer等报道了以染料激光器为光源的光纤分光光度法,用以测定Purex流程中U(Ⅳ),U(Ⅵ),Pu(Ⅲ),Pu(Ⅳ)及Pu(Ⅲ)/Pu(Ⅳ)的混合物。R.A.Malstrom等研究了用416nm激光为激发光源的 光纤荧光光度计,并讨论了用于Savannah河后处理厂铀的在线分析问题。 为探索一种比较经济、简单,又能实现铀的远距离在线分析的测量方法,我们研制了一种以小型脉冲氮分子激光器为激发光源的光纤激光荧光法。我们考虑到Purex流程的实际情况,还研究了硝酸体系中各种因素对铀的荧光强度测量的影响。     

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。     

BESⅢCsI(T1)晶体量能器LED-光纤监测系统的研究

BESⅢCsI(T1)晶体量能器采用LED-光纤监测系统达到监测晶体探测单元和电子学通道的日常运行和性能变化,并在晶体探测单元制造中的存放,组装和运输等过程中通过快速对照测量,检验每个过程中的质量和可靠性.文章介绍LED-光纤监测系统的设计,研究和性能测量.     

EAST中性束注入器电源故障实时监测系统设计

为实现对EAST中性束注入(Neutral Beam Injection,NBI)器电源系统在线运行状态进行实时监测,论文提出了一个基于FPGA、百兆以太网和C#用户终端的全光纤组网实时故障态监测系统解决方案。系统对中性束注入装置各电源的故障态以及输出波形进行实时采集和组网传输,并拥有用户终端进行监测和控制操作,在中性束装置实验运行测试中获得了较好的结果。系统对于大量分散信号的分布采集和远程监测系统设计具有一定参考价值。     

分布式光纤传感技术用于辐射监测的初步探讨

根据光纤受辐照引起缺陷以及着色中心的形成使其光学性能下降产生相应的损耗,提出一种基于分布式光纤传感技术的辐射监测系统的设想,即通过比较输入光和输出光的强度,反演出辐射的剂量,同时利用空分复用技术,采用多个光纤传感器,可以实现对核电站、核武器库等放射物质集中的场所进行大面积多点监测。     

HFETR一回路冷却剂中放射性核素在线监测实验研究

为进行反应堆一回路系统放射性核素在线监测技术研究,开展了对高通量试验堆(HFETR)一回路系统γ放射性核素在线监测实验研究。着重研究了探测器的选择、HPGe探测器的适用性及关键核素选择等问题。通过实验及对实验结果的分析,初步确定了在进行一回路系统γ放射性核素在线监测时,HPGe探测器比闪烁体探测器有明显的优势;初步确认了用HPGe探测器对一回路系统中的γ放射性核素进行在线监测是可行的;初步确定了用于燃料元件破损监测的关键核素:~(138)Cs、~(92)Sr、~(135)I、~(89)Rb、~(134)I、~(142)La、~(133)I、~(138)Xe及~(139)Ba。     

华龙一号LPD在线监测系统误差分析

核电厂采用线功率密度(LPD)在线监测系统对电站运行的实测参数在线计算并显示堆芯线功率密度,能够准确、及时地描述堆芯状态,提高核电厂运行的安全性和经济性。LPD在线监测系统报警限值的设定,需要考虑在线监测系统的总体误差,并留有一定裕量。通过研究华龙一号LPD在线监测系统的总体不确定度的分析方法,将系统的各部分误差通过统计方法综合起来,得到系统的总体误差限值。结果表明,华龙一号采用的LPD在线监测系统误差满足工程要求。     

传感器时间常数随堆测定法

随堆测定传感器时间常数是监测传感器可靠性的一种有效方法.本文以铂电阻温度计为对象,以时序分析为基础,介绍无扰动地测定传感器时间常数的自回归模型法、极大熵谱法和动态数据系统法.这些方法对其他传感器(如热电偶和压力传感器等)时间常数的随堆测定也是有效的.     

On-line monitoring applications in nuclear power plants

The nuclear power industry is working to reduce generation costs by adopting condition-based maintenance strategies and automating testing activities. These developments have stimulated great interest in on-line monitoring (OLM) technologies and new diagnostic and prognostic methods to anticipate, identify, and resolve equipment and process problems and ensure plant safety, efficiency, and immunity to accidents. This paper provides examples of these technologies with particular emphasis on eight key OLM applications: detecting sensing-line blockages, testing the response time of pressure transmitters, monitoring the calibration of pressure transmitters on-line, cross-calibrating temperature sensors in situ, assessing equipment condition, performing predictive maintenance of reactor internals, monitoring fluid flow, and extending the life of neutron detectors. These applications are discussed in the following sections. Emphasis is placed on the principles of a core OLM method - noise analysis - and the technical requirements for an integrated OLM system are summarized.     

Aging management of instrumentation & control sensors in nuclear power plants

Pressure to improve plant efficiency and maximize safety and the increasing age of existing NPPs are forcing the global nuclear power industry to confront the challenges of aging - caused by stressors such as temperature, humidity, radiation, electricity, and vibration - in key instrument & control (I&C) components like pressure transmitters, temperature sensors, neutron detectors, and cables. Traditional aging management methods, such as equipment replacement, required the process to be shut down. Recent aging management technologies, collectively known as online monitoring (OLM), enable plants to monitor the condition and aging of their installed I&C while the plant is operating. Developed through R&D initiatives worldwide, such OLM techniques include low- and high-frequency methods that use existing sensors, such as noise analysis; methods based on test or diagnostic sensors, such as for vibration-measuring accelerometers; and methods, such as the power interrupt (PI) test, based on active measurements made by injecting a test signal into the component under test. A review of these aging management methods, their effectiveness, and their interrelation provides a foundation for understanding the next stage in the evolution of OLM: truly integrated hybrid OLM systems capable of robust condition monitoring in both novel and familiar operating conditions.     

Risk-informed decision-making related to the on-line maintenance

On-line maintenance (OLM) represents the term, which includes testing and maintenance that is performed when the main generator of the nuclear power plant is connected to the grid. OLM on one side helps to decrease the number of activities, which would be performed during the scheduled outage, but on the other side it may contribute to a different level of risk, if the activity is performed when the plant is operating. If the risk of OLM during the power operation is much larger than the risk of similar activity performed during the shutdown the OLM may not be the desired strategy. Additionally, if the risk of OLM during the power operation is not larger than the risk of similar activity performed during the shutdown, the OLM would be the preferred strategy.The objective of the paper is to show risk evaluations of selected OLM activities of selected real plants and to examine if the current practices are suitable for the new nuclear power plants.The results of risk evaluations of OLM and their comparison show that the criteria or guidelines developed for existing plants are not completely suitable for new plants. The new power plants with expected lower risk measures, which can be lower for more than couple of orders of magnitude compared to existing plants, would be able to deal with OLM in plant configurations which would increase the risk for orders of magnitude, but would still be acceptable in terms of risk, if the existing criteria or guidelines are used. Results suggest that the risk guidelines for the OLM should be updated for their use in the case of new nuclear power plants.     

Wireless sensors for predictive maintenance of rotating equipment in research reactors

In 2008–2009, the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) tested the potential of predictive or condition-based maintenance techniques to reduce maintenance costs, minimize the risk of catastrophic failures, and maximize system availability by attaching wireless-based sensors to selected rotating equipment at HFIR. Rotating equipment is an ideal “test case” for the viability of integrated, online predictive maintenance strategies because motors, bearings, and shafts are ubiquitous in nuclear power plants and because the maintenance methods typically performed on rotating equipment today (such as portable or handheld vibration data collection equipment) are highly labor-intensive. The HFIR project achieved all five of its objectives: (1) to identify rotating machinery of the types used in research reactors and determine their operational characteristics, degradation mechanisms, and failure modes, (2) to establish a predictive maintenance program for rotating equipment in research reactors, (3) to identify wireless sensors that are suitable for predictive maintenance of rotating machinery and test them in a laboratory setting, (4) to establish the requirements and procedures to be followed when implementing wireless sensors for predictive maintenance in research reactors, and (5) to develop a conceptual design for a predictive maintenance system for research reactors based on wireless sensors. The project demonstrated that wireless sensors offer an effective method for monitoring key process conditions continuously and remotely, thereby enhancing the safety, reliability, and efficiency of the aging research reactor fleet.     

A Pattern Recognition Method for Nuclear Reactor Core Surveillance

An automatic method for reactor core surveillance is described which employs pattern recognition techniques. A core vibration surveillance method for a pressurized water reactor has been developed which provides the early detection of failures. The method allows on-line application and is able to give quick surveillance decisions (in about one minute). The sensors are eight ex-core ionization chambers. The system is very flexible and adaptable to the statistical properties of the data. The algorithms used are heuristic but need only the definition of a small number of parameters. These parameters may be optimized either by automatic procedures or by explicit experimental methods. Some experimental results are presented.     

Design of copper/carbon-coated fiber Bragg grating acoustic sensor net for integrated health monitoring of nuclear power plant

A nuclear power plant (NPP) is a harsh environment that gives rise to age-related degradation of the plant structures, and eventually leads to radiation leakage that threatens humans. Integrated structural health monitoring (ISHM) technology is a strong candidate for the prevention of the NPP accidents during operation. Prior studies have shown that fiber Bragg gratings (FBGs) and metal-coated fibers have good radiation and high temperature resistance. In this study, a FBG acoustic sensor using a metallic adhesive for installation and a relatively economical copper/carbon (Cu/C)-coated fiber is developed for ISHM of high temperature NPP structures. A chemical method is proposed to remove the Cu/C coating. A 5 mm FBG was successfully inscribed in a Ge-doped silica core through a 7 mm-long silica section with the coating removed. The Cu/C-coated fiber with the same core/clad structure as the standard SMF allowed no-loss fusion splicing, and showed good adaptability to the economical standard fiber, adaptor, connector, and instruments. It showed also good thermal resistance (<345 °C) with no degradation in optical power during the optical transmission. The metallic adhesive used to install the FBG in a one-end-free configuration showed superior bonding reliability during temperature cycles ranging from 25 °C to 345 °C. The FBG reflectivity was stabilized at a 58% drop from the initial reflectivity, and the Cu/C-coated FBG sensor using the metallic adhesive could successfully detect the acousto-ultrasonic waves generated by pencil lead breaking and laser beam excitation.     

Load following control and global stability analysis for PWR core based on multi-model,LQG, IAGA and flexibility idea

The work is to design a nonlinear Pressurized Water Reactor (PWR) core load following control system and analyze the global stability of this system. On the basis of modeling a nonlinear PWR core, linearized models of the core at five power levels are chosen as local models of the core to substitute the nonlinear core model in the global range of power level. The combination control strategy of the Linear Quadratic Gaussian (LQG) control and the Proportional Integral Derivative (PID) control with an optimization process of Improved Adaptive Genetic Algorithm (IAGA) proposed is used to contrive a combined controller with the robustness of a core local model as a local controller of the nonlinear core. Based on the local models and local controllers, the flexibility idea of modeling and control is presented to design a decent controller of the nonlinear core at a random power level. A nonlinear core model and a flexibility controller at a random power level compose a core load following control subsystem. The combination of core load following control subsystems at all power levels is the core load following control system. The global stability theorem is deduced to define that the core load following control system is globally asymptotically stable within the whole range of power level. Finally, the core load following control system is simulated and the simulation results show that the control system is effective.     

An ensemble approach to sensor fault detection and signal reconstruction for nuclear system control

To efficiently control a process, accurate sensor measurements must be provided of the signals used by the controller to decide which actions to actuate in order to maintain the system in the desired conditions. Noisy or faulty sensors must, then, be promptly detected and their signals corrected in order to avoid wrong control decisions. In this work, sensor diagnostics is tackled within an ensemble of Principal Component Analysis (PCA) models whose outcomes are aggregated by means of a local fusion (LF) strategy. The aggregated model thereby obtained is used for both the early detection and identification of faulty sensors, and for correcting their measured values. The fault detection decision logic is based on the Sequential Probability Ratio Test (SPRT). The proposed approach is demonstrated on a simulated case study concerning the pressure and level control in the pressurizer of a Pressurized Water Reactor (PWR). The obtained results show the possibility to achieve an adequate control of the process even when a sensor failure occurs.     

Experiences from sensor tests and evaluation of sensor performance

Sensors are part of the protection system in a nuclear power plant. They are the first link in the chain with influence to the protection system. It is therefore important that the sensors fulfil the demands of reliability and response time. Dynamic investigations of the sensor systems are possible to perform during operation of the plant with the aid of advanced signal analysis. Correctly performed, such an investigation indicates if any sensor with attached sensing lines deviates from the expected dynamic behaviour. The paper presents results from Ringhals 1 in Sweden where snubbers (pulsations dampers with exchangeable needles) were used in the sensing lines to differential pressure sensors. The dynamic influence of the needles was equivalent to a time constant = 0.55 s. It was clear that the response time was longer than the demanded response time with respect to the sensors task in the safety system. By eliminating the needles the demand on the response time was fulfilled. The deviation lead to a report to the Swedish Nuclear Power Inspectorate (equivalent to a Licensee Event Report) with INES level 1. The title of the report is “Too long response time caused by not correct installed pulsation dampers in swelling transmitters”. The paper proves convincingly that signal analysis of multiple sensor signals is a fruitful way to observe dynamic deviations in measurement systems in nuclear power plants.