The newest two-phase flow control devices in LWR equipment based on ultrasonic and WAT-technology

This paper deals with new acoustic methods of two-phase flow diagnostics used to carry out research in the fields of nuclear power thermophysics and nuclear power plant (NPP) technologies equipment control. All the designs are to be used under extreme conditions, characteristic for water coolant, with temperature up to 350°C and pressure 20 MPa. All the safety and reliability requirements are met. The methods use waveguide ultrasonic transducers for longitudinal and bending waves, made according to specially designed technology (waveguide acoustic transducers — WAT technology). This paper deals with the operating principles of transducers and processing device physical models as well as some results on the practical use of this equipment. The method of acoustic impedance is based on measuring attenuation of a longitudinal or bending ultrasonic wave in a thin-walled tube diameter vapour fraction or the level of the coolant in the tank. The waveguide transducers, designed by the centre, use bending waves of a surface type. They enable us to carry out diagnostics of the liquid film on the inner surface of the tube or discover gas inclusions in the liquid flow. The paper touches upon the method of acoustic emission for measuring moisture content in a steam flow.     

Monitoring the filling level of the rpv using stationary emus-probes

The determination of the filling level and the initiation of growth of bubbles in vessels and pipes containing fluids is an essential component of monitoring during operation.The ultrasonic pulse-echo-method is a measuring procedure suited for this purpose and applicable from the outside. Piezoelectric ultrasonic transducers can be used in principle at a temperature of 300°C, but in practice these transducers are not preferred because of the expense and inconvenience in coupling them to the vessel wall.These problems are solved using lectro agnetic- ltrasonic-(EMUS-)transducers. Due to physical reasons a longitudinal wave is generated in the fluid by the refraction of a shear wave in the vessel-wall. The filling level is measured in a pitch and catch-technique by a mirror reflection of the longitudinal wave at a construction element inside the vessel. This paper reports on laboratory investigations concerning the applicability of the technique and first experiences with an EMUS-prototype system installed in a nuclear power plant.     

核电厂蒸汽发生器与主泵对接焊缝检查系统开发

AP1000核电厂蒸汽发生器出口接管与主泵泵壳对接焊缝泵壳侧为粗晶奥氏体铸造材料,由于该焊缝壁厚大、超声衰减、晶粒散射严重等导致焊缝的超声检测技术开发难度大。本研究采用特殊的设计,开发了一套从蒸汽发生器出口接管内壁实施超声检测的自动检查系统,并将该系统应用于国内某AP1000核电厂的役前检查。结果表明,该检查系统完全满足现场检查要求,检验结果与焊缝出厂检验结果具有良好的一致性。      

AP1000核电厂主泵法兰螺栓在役检查系统开发

AP1000核电厂反应堆主泵法兰螺栓是在役检查重要监督项目之一,目前国内尚无针对该部件的在役检查系统及应用案例。本文结合AP1000主泵法兰螺栓结构特点、现场高剂量环境及复杂检查条件分析,设计开发了一套从螺栓中心孔内壁实施超声检测、适用于在役检查要求的主泵法兰螺栓在役超声检查系统。主泵模拟体上的调试试验结果表明,该系统可实现周向运行、垂直方向避障、专用超声探头与螺栓孔精确对中调节等功能,进而实现对主泵法兰螺栓的超声扫查。工程应用结果证明本系统满足AP1000核电厂主泵法兰螺栓在役检查现场要求,具有较高的可靠性和良好的适用性。      

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

On the numerical assessment of the thermo-mechanical performances of the DEMO Helium-Cooled Pebble Bed breeding blanket module

Within the framework of the European DEMO Breeder Blanket Programme, a research campaign has been launched by University of Palermo, ENEA-Brasimone and Karlsruhe Institute of Technology to theoretically investigate the thermo-mechanical behavior of the Helium-Cooled Pebble Bed (HCPB) breeding blanket module of the DEMO1 blanket vertical segment, under normal operation and over-pressurization loading scenarios.The research campaign has been carried out following a theoretical–computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. A realistic 3D FEM model of the HCPB blanket module central poloidal–radial region has been developed, including one breeder cell in the toroidal direction and all the five cells in the poloidal one. No Breeder Units have been modeled, their presence being simulated by effective thermo-mechanical loads.Two sets of uncoupled steady state thermo-mechanical analyses have been carried out with reference to the investigated loading scenarios. In particular, under normal operation scenario (level A) the module has been supposed to undergo both 8 MPa coolant pressure on its cooling channel walls and thermal deformations due to the flat-top plasma operational state thermal field, while under over-pressurization scenario (level D) it has been assumed to experience 8 MPa coolant pressure on its internal walls while operating at normal operation thermal level. Results obtained are presented and critically discussed according to the SDC IC code.     

Experimental and numerical evaluation of IR thermography method for Final Acceptance Tests of the ITER divertor dome

The divertor dome (DO), being part of the ITER divertor, is designed to extract the major part of the plasma thermal energy. As a plasma-facing component (PFC), the DO experiences high heat fluxes (up to 5.0 MW/m²). Such severe operation conditions of the DO imply stringent requirements for the DO design and its cooling system to ensure the required temperature operation regime of the dome. Hence, Final Acceptance Tests (FAT) shall be performed on each DO final assembled component with the aim to demonstrate that none of parallel coolant channels are completely or partially blocked. The paper presents the results of the analytical and experimental testing of the thermography method capability to perform the FAT. The aim is to determine defective hypervapotrons of the divertor dome. The method consists in contactless measurement of the dynamic temperature field of the PFC surface at a step-like increase (from zero to constant value) in the coolant flow rate with a temperature higher than that of the hypervapotron.     

Effect of cladding on the initiation behaviour of finite length cracks in an RPV under thermal shock

An analytical method has been developed and verified by three-dimensional elastic-plastic finite element analyses to evaluate stress intensity factors for finite length through clad and subclad cracks in reactor pressure vessels (RPV) under loss of coolant accident conditions. The method is applied to thermohydraulic transients of the RPV of KKS. The results demonstrate the margin of safety for the RPV for end of life material conditions.     

A New Non-destructive Technique for Hydrogen Level Assessment in Zirconium Alloys using EMAR Method

In order to study the applicability of EMAR (electromagnetic acoustic resonance) method to non-destructive hydrogen level assessment in fuel spacer bands at pool side, an ultrasonic transmitter and receiver together with an EMAT (electromagnetic transducer) were used. Unirradiated Zircaloy-2 thin plates were hydrogen charged for the measurements. An irradiated fuel cladding tube was also used to examine the detection sensitivity of the resonance spectrum of the irradiated material. The following results were obtained. Acoustic anisotropy Δf, defined by using two resonance frequencies for shear waves with different polarization, was adopted as a parameter to express the ultrasonic resonance property. A hydrogen concentration dependence of Δf was observed in the range up to 1,200 ppm. Specimen thickness and oxide thickness were found to have negligible effect, on Δf, and liftoff of the sensor up to 1mm did not affect the Δf value. The acoustic anisotropy proposed in this paper was not sensitive to any of specimen dimension, surface condition, or sensor liftoff.     

压水堆核电站常规运行工况下气载放射性物质释放源项的计算

本文介绍一个自行编制的用于计算压水堆核电站在常规运行工况下气载放射性物质向环境释放量的计算机程序ＭＧＡＬＥＳ。采用ＯＲＩＧＥＮ２程序，根据燃料元件的成份和燃耗情况计算堆芯的放射性核素谱；用放射性物质经堆芯向一回路迁移的逃脱率系数计算一回路冷却剂中的放射性核素浓度；再考虑核电站实际运行过程中一、二回路冷却剂的泄漏以及通风、除气等过程，计算其正常运行工况下气载放射性物质向环境的释放量。     

Ultrasonic study of UO2: effects of porosity and grain size on ultrasonic attenuation and velocities

Ultrasonic techniques applied to nuclear fuel characterisation are developed in our group since 1996. Before applying our methods to irradiated fuel, we are searching sensitive parameters which could give interesting information. That is the reason why only results concerning non-irradiated UO₂ are presented. This paper mainly deals with the investigation of a relevant acoustic parameter: the attenuation. Indeed, the ultrasonic attenuation in UO₂ as a function of the operating ultrasonic frequency has been measured on samples with various microstructures: variable fraction volume porosity (1–6%) and grain size (10–90 μm). Using a 15 MHz operating frequency, no attenuation has been observed. With frequencies around 40 MHz, we show that the measured ultrasonic attenuation is only sensitive to grain size (no effect of porosity has been observed). On the contrary, the ultrasonic velocities (which are very sensitive to porosity) are not affected by the sizes of the grains. These reversed and non-correlated effects constitute an interesting tool for UO₂ study because two aspects of the microstructure can be studied separately with ultrasonic waves.     

Measurements of the acoustic field on austenitic welds: a way to higher reliability in ultrasonic tests

In nuclear power plants many of the welds in austenitic tubes have to be inspected by means of ultrasonic techniques. If component-identical test pieces are available, they are used to qualify the ultrasonic test technology. Acoustic field measurements on such test blocks give information whether the beam of the ultrasonic transducer reaches all critical parts of the weld region and which transducer type is best suited. Acoustic fields have been measured at a bimetallic, a V-shaped and a narrow gap weld in test pieces of wall thickness 33, 25 and 17 mm, respectively. Compression wave transducers 45, 60 and 70° and 45° shear wave transducers have been included in the investigation. The results are presented: (1) as acoustic C-scans for one definite probe position, (2) as series of C-scans for the probe moving on a track perpendicular to the weld, (3) as scan along the weld and (4) as effective beam profile. The influence of the scanning electrodynamic probe is also discussed.     

变流量工况下^16N监测反应堆功率的研究

本文从理论上对变流量工况下利用＾１６Ｎ监测压水堆功率进行了分析，推导出反应堆－回路中＾１６Ｎ放射性强度与冷却剂流量之间的关系，得出变流量工况下＾１６Ｎ测量反应堆功率的简化公式，提出了变流量工况下＾１６Ｎ监测反应堆功率的方法，并报告了应用该方法在ＨＦＥＴＲ（高通量工程试验反应堆）上的研究及试验结果。     

Prototype EMAT system for tube inspection with guided ultrasonic waves

A prototype system with full computer support for ultrasonic inspection of ferritic tubes using guided waves is described. The ultrasonic waves are launched and received with the aid of electromagnetic acoustic transducers which are layed out as linear phased arrays. The array structure provides a good axial directivity for the transducers so that the probe can be positioned anywhere along the tube length sequentially transmitting ultrasonic pulses in the foreward and backward directions. While the probe is fixed at one axial position during inspection the tube length is measured by the system and flaws are detected from returning ultrasonic echos. Results of the inspection of tubes with natural flaws are given and the wavelength-spectrum of the ultrasonic mode used for the inspection is discussed with respect to flaw depth sizing.     

核电厂声学泄漏监测系统的设计和验证

介绍VVER-1000型核电厂声学泄漏监测系统的设计基准和功能,给出判断泄漏过程、确定泄漏量和泄漏位置的系统算法。建立主回路声模型结构图,计算得到环路背景噪声分布,并和实际机组上的试验结果进行对比。建立管道模型的试验台架,并对管道模型进行了试验验证。根据试验数据得到用于计算泄漏量大小和泄漏位置的相关系数。通过核电厂声学泄漏监测系统的设计和验证,为开发田湾核电厂1、2号机组在线的核电厂声学泄漏监测系统奠定了基础。      

Heat transfer from a partially uncovered reactor core

This paper documents a model which has been developed for predicting the temperature distribution along a “flow channel” of a pressurized water reactor during simulated, uncovered core conditions. In the model, heat conduction along the fuel element, convection from the surface to the coolant, radiation exchange between the clad surface and steam, and surface exchange between adjacent fuel rods are considered. Variations of the thermophysical properties of the fuel road and of the coolant with temperature are accounted for, but oxidation of Zircaloy is not modeled. Extensive sensitivity studies on the effects of heat generation in the core, steam velocity, pressure level, uncovered core height, presence of hydrogen gas in the coolant, power skew, clad emissivity, and convective heat transfer correlations have been examined. The results show that the importance of radiation in comparison with convection increases with an increase in the fuel rod temperature, pressure, and clad emissivity.     

Effects of Coolant Flow on Light Water Reactor Fuel Behaviors during Reactivity Initiated Accident

This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.     

秦山核电厂调试期间一回路的水质控制

为尽量减少秦山核电厂一回路系统设备的腐蚀,以及减少冷却剂循环污染和降低一回路系统的放射性,在冷态试验和热态功能试验阶段,堆装料、启动和功率运行试验阶段,都进行了一回路水质控制。本文介绍了一回路水质控制方法,并着重介绍了冷却剂的pH值、氯离子、溶解氧、溶解氢的控制以及硼与锂的协调。秦山核电厂的调试获得令人满意的成功,一回路水质控制也是卓有成效的。     

History of the engineering and scientific-technical contribution of BR-5 and BR-10 reactors to the development of fast sodium-cooled reactors

The first experimental sodium-cooled reactor BR-5 in Europe was built and put into operation in a record short time – three years (1956–1959). The main goal of building such a reactor was to master the use of sodium coolant and sodium equipment under radiation-hazardous operating conditions. It is shown that the reactor made it possible to solve other problems also. The first sodium apparatus, systems for monitoring and purifying the coolant, and methods and facilities which were later used in the development of fast power reactors were mastered on BR-5. The components of successful multiyear (43 years) operation and the results of investigations are presented. The article focuses primarily on the experiential aspects which were later used to develop and operate subsequent fast reactors. Translated from Atomnaya énergiya, Vol. 106, No. 3, pp. 134–140, March, 2009.     

LOCA工况下环形燃料元件内包壳外压屈曲试验研究

环形燃料一种安全高效的新型核燃料。为对环形燃料元件冷却剂丧失事故（LOCA）下整体受压失效形式的问题进行研究,将环形电加热棒、模拟芯块和试验件组装成试验装置,在空气环境中,以环形电加热棒外加热的方式,对环形燃料元件内包壳进行了外压屈曲试验,并将试验屈曲压力与Bresse?Bryan公式计算结果和特征值屈曲数值模拟分析结果进行了对比分析。结果表明：Bresse?Bryan公式计算结果除以安全系数m=2?5得到的结果高于试验结果而不够保守,试验结果分布于特征值屈曲数值模拟分析结果的1/5?1/3之间。本文结果可为环形燃料元件安全评价及后续工程化提供基础数据。