Determination of thermocouple transfer-functions and fluid-flow velocities by temperature-noise measurements in liquid sodium

The time response of TI-, TM- and intrinsic thermocouples has been investigated in sodium by determination of frequency dependent thermocouple transfer functions and related delay times. Experimentally temperature fluctuations (temperature noise), generated in sodium by heated channels or injection of hot sodium, have been used as driving sources. The measured transfer functions and delay times have been compared to predictions from theoretical models. Good agreement was found. Intrinsic and TM-type thermocouples are best suited if fast response is required. Delay times in the range of one millisecond or upper break frequencies of 1000 Hz were determined for intrinsic thermocouples.

The determination of sodium flow velocities by the transit time correlation technique showed systematic error sources for analysis at low frequencies due to intermediate heat storage between the thermocouple positions. Statistical errors have been determined as function of thermocouple distance, frequency range and time of analysis. Minimum errors are obtained for thermocouple distances from 100 to 150 mm and in the frequency range 90 to 120 Hz. A transit time correlator, which automatically tracks the maximum of the cross-correlation function and displays the velocity, has been successfully used for signals with correlation coefficients greater than 0.3.     

Integrating cross-correlation techniques and neural networks for feedwater flow measurement

This paper reports an early progress of a feasibility study of a computational intelligence approach to the enhancement of the accuracy of flow measurements in the framework of an ongoing cooperation between Tecnatom s.a. in Madrid and the OECD Halden Reactor Project (HRP) in Halden. The aim of this research project is to contribute to the development and validation of a flow sensor in a nuclear power plant (NPP). The basic idea is to combine the use of applied computational intelligence approaches (noise analysis, neural networks, fuzzy systems, wavelets etc.) with existing traditional flow measurements, and in particular with cross-correlation flowmeter concepts.     

Flow measurements by noise analysis of thermocouple signals from the BOCA experiment at the Studsvik R2 reactor

At the Studsvik research reactor R2, a Boiling Capsule (BOCA) is used for long-term irradiation of BWR and PWR fuel rods. The BOCA experiment consists of a pressurised container that can hold a number of fuel rods in a bundle type configuration. The water flow inside the tube is driven by natural circulation. The coolant flow rate is not normally measured in the BOCA rig. Only thermocouples, measuring the water temperature at pertinent locations, are located inside the pressure tube. To confirm calculated values of the flow rate, transit time determination through the cross-correlation technique has been implemented.

Campaigns of noise measurements have been performed at five different occasions. The measurement campaigns have included 10 thermocouples at 3–4 different power levels. The results for the flow rate range between 0.15 and 0.35 m/s depending on reactor power level. The statistical accuracy of the results has also been evaluated. This paper shows that signal processing of thermocouple signals can be used to obtain rather accurate values of the flow rate in BOCA.     

Evaluation Method of Statistical Error in. Transit Time Measured with Cross-Correlation of Two Detector Signals

In the transit time method for measuring flow velocity, the cross-correlation function of two detector signals is computed; the delay time which maximizes the function gives the transit time of the fluid between the detectors. The cross-correlation function is mathematically defined for an infinite observation time, however actually it is computed based on a finite length of time. Due to this approximation, the measured cross-correlation function has a statistical uncertainty, which induces a scattering on the measured transit time.

In this paper, it is shown that the statistical error of the transit time may be expressed by using either the derivatives of correlation functions or the power spectral densities. The spectral expression has a merit of giving an intuitive understanding of the nature of the error and it can be seen that the error becomes smaller as the higher frequency components of the spectra increase. The error also becomes smaller when the coherence function has a larger value and/or the observation time extends for a longer period.

A simplified evaluation method of the error is proposed using a trapezoidal approximation for the power spectra and this method is applied to the transit time measurement by electro- magnetic-flowmeters. The evaluated statistical error closely agreed with the measured scattering.     

华龙一号小幅功率提升研究

对华龙一号热功率精度进行了分析,计算了蒸汽发生器出口压力测量精度、给水温度测量精度和给水流量测量精度对华龙一号热功率精度的贡献度,通过定量化的数据证明了主给水流量测量精度对热功率计算精度的影响最大。基于目前孔板流量计精度低,长期使用精度劣化的问题,提出采用高精度（0.3%）的超声波流量计来测量主给水流量,计算结果表明,采用超声波流量计可以获得0.97%的功率提升。      

Actual flow calibration of a feedwater flowmeter using a high Reynolds number facility at NMIJ

The results of calibration tests of the feedwater flowrate of ultrasonic flowmeters used in a nuclear power plant for variety of upstream conditions obtained using the new high Reynolds number calibration facility at NMIJ are described. In this examination, the measurements are performed for five pattern pipe layouts with one or two elbows. The flow conditioners installed upstream of the flowmeter are the tube bundle type and the Mitsubishi, which are normally used in nuclear power plants. The calibration result for each flowmeter are largely different for each flow conditioner and each upstream pipe layout, except in some special cases. Moreover, the trend of the correction factor with Reynolds number is not uniform for each case. Furthermore, some differences were observed for individual flowmeters. It is recommended that the feedwater flowmeter, especially when used to perform measurement uncertainty recapture, is calibrated based on the actual pipe layout and the Reynolds number corresponding to the actual nuclear power plant conditions.     

Measurement of the coolant flow rate using correlation of temperature fluctuations

The transit time of the coolant, and thus its velocity, has been measured using the temperature fluctuation at the outlet of a reactor core. An impulse response function estimation is introduced, which substitutes the widely used cross-correlation measurement technique. It is shown in theory and practice that the time delay estimation is improved when using the impulse response function instead of the cross-correlation function in parameter estimation. Extremely low velocities (down to 2 cm/sec) have been measured in a natural circulation regime in a research reactor.     

Investigation of the two-phase flow in a boiling water reactor using neutron-noise technique

In this paper, the fluctuations of the neutron flux (“neutron noise”) of the Mühleberg BWR are investigated. Above 2 Hz, the noise measured by the in-core neutron detectors is driven exclusively by local fluctuations of the void fraction. Characteristic changes of the neutron-noise signature along the axis can be attributed to changes of flow pattern. By measuring the phase lag between pairs of axially placed neutron detectors, the transit time of the steam between the detectors can be evaluated. The measured transit times are applied to the study of two-phase flow in the core. The neutron-noise method has the advantage of providing in-core information under operational conditions.     

Numerical evaluation of SmCo permanent magnet flowmeter measuring sodium flow in a low flow rate regime using FLUENT/MHD module

One of the several key subsystems in the test facility of Korean sodium-cooled fast reactor is a plugging meter system, which measures the impurities in the sodium using an indirect online technique. To measure the low flow rate, a permanent magnet flowmeter was developed owing to its inherent fast response time, non-invasive characteristic, relatively accurate flow rate measurements, and excellent linearity between the flow rate and flowmeter signal. However, several limitations have been reported in the experimental evaluation of the flowmeter under low flow rate conditions given the measurement capability of the current experimental facility. Thus, the performance of the flowmeter was evaluated numerically using a commercial computational fluid dynamics tool, a FLUENT/MHD module, based on the finite volume method with the help of electromagnetic analysis software, ANSYS MAXWELL. The FLUENT/MHD module was validated by comparing the simulation results with the experimental results. The relative error of the FLUENT simulation was estimated to be approximately 0.24% compared with the experimental results. After the validation process, MHD simulations were conducted to calculate the flowmeter voltage signals versus flow rates, especially in a low flow rate regime, where the linearity between the flow rate and flowmeter signal was carefully analyzed.     

Measurement System of Bubbly Flow Using Ultrasonic Velocity Profile Monitor and Video Data Processing Unit, (II)

The authors have developed a measurement system which is composed of an ultrasonic velocity profile monitor and a video data processing unit in order to clarify its multi-dimensional flow characteristics in bubbly flows and to offer a data base to validate numerical codes for multi-dimensional two-phase flow. In this paper, the measurement system was applied for bubbly countercurrent flows in a vertical rectangular channel. At first, both bubble and water velocity profiles and void fraction profiles in the channel were investigated statistically. Next, turbulence intensity in a continuous liquid phase was defined as a standard deviation of velocity fluctuation, and the two-phase multiplier profile of turbulence intensity in the channel was clarified as a ratio of the standard deviation of flow fluctuation in a bubbly countercurrent flow to that in a water single phase flow. Finally, the distribution parameter and drift velocity used in the drift flux model for bubbly countercurrent flows were calculated from the obtained velocity profiles of both phases and void fraction profile, and were compared with the correlation proposed for bubbly countercurrent flows.     

Effect of low-frequency ultrasound on flow rate measurements using the ultrasonic velocity profile method

This study presents a low-frequency ultrasonic propagation analysis using the finite-element method (FEM). Experimental results of flow rate measurements using the ultrasonic velocity profile (UVP) method are also presented. The ultrasound frequency, pipe diameter, and pipe wall thickness are 0.274 MHz, 590.6 mm, and 9.5 mm, respectively. Six waves are generated per ultrasound pulse. To analyze the entire pipe region, the FEM is combined with the Kirchhoff method. The experiments of flow rate measurements are conducted using the high Reynolds number calibration facility at the National Metrology Institute of Japan. The range of the Reynolds number is from 4.4×10⁶ to 1.7×10⁷. Wide spreading of the ultrasonic beam in the axial direction of the pipe is observed because of multiple reflections in the pipe wall. This wide beam affects the measured velocity profile, particularly in the region near the pipe wall. In addition, the flow rate errors are approximately 10% (deviating by 1.1%) across the investigated range of Reynolds number. This result suggests that the experimental flow rate errors might be used as correction factors of flow rate measurements using the UVP method.     

CANDU 300: The economic small reactor

The CANDU nuclear power system has evolved in a carefully planned and systematic manner over the past 40 years. Thirty-one CANDU power stations are now in commercial operation, or under construction, world wide.Changes in the demands of the world power market, and growing interest in smaller nuclear units led AECL to develop the CANDU 300, which has a net electrical output in the range of 450 MW.AECL initiated design work on the CANDU 300 in 1982. The substantial effort devoted to this program over the past five years has produced a “small” CANDU power station that is economically competitive with both large nuclear and coal fired generating stations.The CANDU 300 makes substantial advances in the areas of station layout, constructability, maintainability, and operability and plant life extension capability while utilizing proven systems, components and concepts. All key components, including steam generators, pressure tubes, fuelling machine, fuel, and coolant pumps are essentially identical to those in service in the very successful CANDU 600 stations.This paper provides an overview of the CANDU 300 with emphasis on factors impacting economics and performance.     

CANDU technology for Generation III+ and IV reactors

Atomic Energy of Canada Limited (AECL) is the original developer of the CANDU^® reactor, one of the three major commercial power reactor designs now used throughout the world. For over 60 years, AECL has continued to evolve the CANDU design from the CANDU prototypes in the 1950s and 1960s through to the second-generation reactors now in operation, including the Generation II+ CANDU 6. The next phase of this evolution, the Generation III+ Advanced CANDU Reactor™ (ACR™), continues the strategy of basing next generation technology on existing CANDU reactors. Beyond the ACR, AECL is developing the Generation IV CANDU Super Critical Water Reactor.Owing to the evolutionary nature of these advanced reactors, advanced technology from the development programs is also being applied to operating CANDU plants, for both refurbishments and upgrading of existing systems and components. In addition, AECL is developing advanced technology that covers the entire life cycle of the CANDU plant, including waste management and decommissioning. Thus, AECL maintains state-of-the-art expertise and technology to support both operating and future CANDU plants.This paper outlines the scale of the current core knowledge base that is the foundation for advancement and support of CANDU technology. The knowledge base includes advancements in materials, fuel, safety, plant operations, components and systems, environmental technology, waste management, and construction. Our approach in each of these areas is to develop the underlying science, carry out integrated engineering scale tests, and perform large-scale demonstration testing. AECL has comprehensive R&D and engineering development programs to cover all of these elements.The paper will show how the ongoing expansion of the CANDU knowledge base has led to the development of the Advanced CANDU Reactor. The ACR is a Generation III+ reactor with substantially reduced costs, faster construction, and enhanced passive safety and operating margins. The ACR is also the basis for the Generation IV Super Critical Water Reactor, which extends operation to higher temperatures and pressures.     

Thermal power estimation by fouling phenomena compensation using wavelet and principal component analysis

A small percentage of reactor thermal power can be overestimated because of fouling phenomena in a secondary feedwater flowmeter. This study proposes a signal processing technique for the compensation of a degraded flowmeter such a secondary feedwater flowmeter in nuclear power plants. The technique proposed is mainly focused on noise classification and step-by-step noise reduction. The noises focused are classified into the rapid distortion caused by environmental interference, the flow fluctuation according to plant state transition and the degradation by fouling phenomena qualitatively. The multi-step de-noising technique reduces each noise by three techniques step-by-step. The wavelet analysis as a low frequency pass filter to remove the rapid distortion, the linear principal component analysis (PCA) to predict a steady-state value from the fluctuation, and the non-linear PCA implemented as an autoassociative neural network (AANN) to predict an original value from the signal including fouling phenomena are developed. The main purpose of this approach is to make an AANN concentrate on compensating the degradation by fouling phenomena itself. For the demonstration the signals from a simulator and signal modeling were used so that the role and the performance of each noise removal step was represented. In addition a thermal power deviation estimator is proposed to recognize the degradation effect of each operating parameter for reactor thermal power calculation.     

Simultaneous measurement of liquid velocity and interface profiles of horizontal duct wavy flow by ultrasonic velocity profile meter

A simultaneous measurement of the liquid velocity and interface profiles was performed for stratified-smooth and wavy flows in a horizontal duct using a ultrasonic velocity profile (UVP) meter. The influences of the reflections of ultrasonic pulses at the gas–liquid interface and channel bottom were reduced by using an absorbent for the ultrasonic pulses on the duct bottom wall and optimization of the liquid level and time interval between pulses. For a smooth–stratified flow, good comparison was obtained with a velocity profile obtained by particle tracking velocimetry (PTV) for video pictures taken simultaneously at the UVP measurement. Polystyrene beads were used as the reflector and tracers respectively, for the UVP and PTV measurements. The velocity profiles measured for a wavy flow with periodically-generated interfacial waves agreed well with the theoretical prediction for solitary waves. Turbulence component appeared in the velocity profiles of both the smooth–stratified and wavy flows.     

基于CFD分析的文丘里流量计设计优化

质量流量是核电站热功率核算的关键参数之一,核电站一般采用文丘里流量计和孔板流量计同时测量,然而在低流量区文丘里流量计呈现出明显波动,其参数不稳定严重影响核电站的正常运行。本文基于理论分析结合数值分析,发现脉动流动是导致文丘里流量计测量波动的主要原因。基于分析结果,对文丘里流量计提出了优化设计方案,通过在文丘里管上游集成流量调整装置,从而减小脉动流,有效提升文丘里流量计的稳定性。此外,开展了集成不同类型流量调整装置的文丘里流量计压损特性数值研究,结果表明K-Lab型流量调整装置阻力较小。本文提出的方案可有效提升文丘里流量计测量精度。     

Results of the temperature noise measurements in subchannels of sodium cooled electrically heated 28 rod bundle behind a blockage

For LMFBR safety studies a 28 rod bundle has been built at Petten (cooperation of GfK and ECN), representing a 60-degrees section of an SNR-300 fuel element having a 70% flat type central blockage. The aims of the temperature noise measurements were to determine the subchannel coolant velocities behind the blockage to study the mixing of coolant in subchannels of different temperature from behind the blockage to the outlet and to study the temperature noise due to boiling in a subchannel. The temperature noise measurements were carried out in parallel to the other measurements (temperature distribution, etc.), using signals of fourteen subchannel thermocouples placed in five measuring planes behind the blockage. The single phase measurements were made with several heat fluxes (5 W/cm² to 120 W/cm²), inlet flows (0.25 to 3 m/s) and inlet temperatures (250°C to 600°C). Two phase flow is initiated and sustained either by a slow and continuous pressure reduction or by stepwise reduction of the main flow. The temperature noise signals were amplified and recorded in analog form. Later the signals were digitized and analysed by digital computers. Part of the signals was also processed by a hardware correlator. The experimental results of the temperature noise measurements will be shown for the different conditions of the loop. Measurements clearly show the following effects:

• - the recirculation flow pattern due to the vortex in the wake behind the blockage;

• - the dependence of r.m.s. value of the noise on the heat flux and the coolant flow;

• - the increase in noise and change in power spectra when going from single phase to the boiling condition.