Application of Neutron Radiography to Visualization and Void Fraction Measurement of Air-Water Two-Phase Flow in a Small Diameter Tube

Abstract

The purpose of this study is to investigate the feasibility of visualization and void fraction measurement of air-water two-phase flow in a small diameter tube (I.D.: 4.08mm) by using the real-time neutron radiography and image processing techniques. Video images of two-phase flow were taken by using the real-time neutron radiography system (thermal neutron radiography facility No. 2) installed at the Japan Research Reactor 3 M of the Japan Atomic Energy Research Institute. The shape of bubbles and its moving behavior were clearly observed from the video images. The image corrections for dark current, shading, field intensity fluctuation and electrical system drift were examined in order to measure the void fraction from the video images. Though, generally speaking, the effect of the scattered neutron could not be ignored for quantification of the images taken by the neutron radiography, the scattered neutron could not affect the final results of void fraction in the case of a small diameter tube. The void fraction calculated from the corrected images was correlated well with the drift flux equation, indicating that the existing drift flux equation could be applied to predict the void fraction of two-phase flow in a small diameter tube. It was demonstrated that the real-time neutron radiography technique could be useful for measuring the void fraction of two-phase flow in a small diameter tube.     

Visualization and measurement of two-phase flow by using neutron radiography

To apply neutron radiography (NR) technique to fluid research, high frame-rate NR with a steady thermal neutron beam has been developed in the present research program by assembling up-to-date technologies for neutron source, scintillator, high-speed video and image intensifier. This imaging system has many advantages such as a long recording time (up to 21 min), high-frame-rate (up to 1000 frames s⁻¹) imaging and no need for triggering signal. Visualization of air-water two-phase flow in a metallic duct was performed at the recording speeds of 250, 500 and 1000 frames s⁻¹. The qualities of those consecutive images were good enough to observe and measure the flow structure and the characteristics. It was demonstrated also that some characteristics of two-phase flow could be measured by using the present imaging system. Image processing technique enabled measurements of various flow characteristics in two ways. By utilizing geometrical information extracted from NR images, data on flow regime, rising velocity of bubbles, and wave height and interfacial area in annular flow were obtained. By utilizing attenuation characteristics of neutrons in materials, measurements of void profile and average void fraction were performed. It was confirmed that this new technique may have significant advantages both in visualizing and measuring high-speed fluid phenomena when the other methods such as an optical method and X-ray radiography cannot be applicable.     

Three-dimensional visualization of void fraction distribution in steady two-phase flow by thermal neutron radiography

Three-dimensional void fraction distributions of a steady air–water two-phase flow in a 4×4 rod-bundle with circular ferrule type spacers were measured by neutron radiography using a CT method. The high flux thermal neutron radiography system at JRR-3M in JAERI was used. Two-phase flow was visualized with a SIT tube camera and time-averaged one-dimensional cross sectional averaged void fraction distributions were calculated. Visualization with high spatial resolution up to 0.18 mm was carried out by using a cooled CCD camera. Projections in 250 directions were obtained and were reconstructed by a filtered back projection method after using some image processing techniques. Animations were made to show the three-dimensional distributions. One-dimensional and three-dimensional void fraction distributions of the steady state two-phase flow in the rod bundle near the spacer were clearly visualized.     

中子照相∑定量测量方法

本文描述了一种利用中子照相进行定量分析的方法,介绍了该方法的基本原理,并应用粒子输运程序MCNP模拟分析样品内部散射中子随样品与探测器之间距离变化对中子成像的影响.为削弱散射中子对中子透射成像图像信息的影响,对穿透样品的厚度进行了定量分析.     

The Void and Hole Effect in D2 Moderated Reactor

The reactivity cbange due to increase in the radius of empty hole was measured in a D₂O moderated reactor and some results differing from experiments with ZEEP were obtained. It can be concluded that the streaming in a hole is not so effective for reactivity. In measuring neutron flux in a void, a flat thermal neutron flux distribution was obtained and it has been concluded that the neutrons leaking through the empty hole or the void do not consist of thermal neutrons but fast neutrons for the most part. The experimental result of reactivity change due to the void location in the core indicates that the relation between the void location and the reactivity change is independent of the neutron flux distribution.     

石墨双晶单色器性能检验及应用

中子能量选择成像作为一种前沿中子成像技术,可实现传统白光中子照相技术无法实现的功能,如研究工程材料中的晶粒分布、应变/应力分布、织构测量和相变分析等。本文依托中国先进研究堆(CARR)上的中子成像测试平台,在国内首次研制了石墨双晶单色器,建立能量选择中子成像技术。飞行时间实验测试结果表明,该双晶石墨单色器在选择4×10^-10 m的中子时波长分辨率可达2.6%,优于3.0%的设计指标。虽然在冷源未开启的实验条件下该石墨双晶单色器产生了较多的次级中子,但基于现有条件开展的镍基高温合金的中子能量选择成像实验能清楚分辨特定取向微晶粒的形貌及分布。结果表明研制的石墨双晶单色器可在CARR上开展能量选择成像实验,随着未来CARR冷源的开启,次级中子数量降低,中子束流品质进一步提高,将开展高质量的中子能量选择成像实验。     

High-speed neutron radiography for monitoring the water absorption by capillarity in porous materials

Fluid flow through porous natural building stones is of great importance when studying their weathering processes. Many traditional experiments based on mass changes are available for studying liquid transport in porous stones, such as the determination of the water absorption coefficient by capillarity. Because thermal neutrons experience a strong attenuation by hydrogen, neutron radiography is a suitable technique for the study of water absorption by capillarity in porous stones. However, image contrast can be impaired because hydrogen mainly scatters neutrons rather than absorbing them, resulting in a blurred image. Capillarity results obtained by neutron radiography and by the European Standard 1925 for the determination of the water absorption coefficient by capillarity for natural building stones with a variable porosity were compared. It is illustrated that high-speed neutron radiography can be a useful research tool for the visualization of internal fluid flow inside inorganic building materials such as limestones and sandstones.     

Measurement of void fraction in bubbly-slug flow with a constant electric current method

In several void fraction measurement methods, a constant electric current method which is one of conductance methods is focused in the present study. By using this method, void fraction can be measured with higher temporal resolution. However, it has been mainly applied to annular flow in previous studies. In the present study, Maxwell's estimation, Bruggemann's estimation, low void fraction approximation and new estimations which consider the bubble shape are applied in order to measure more accurately void fraction of dispersed bubbly flow and slug flow. To understand the effect of bubble shapes and flow patterns, void fraction was measured by the constant electric current method for a rising single spherical bubble and a rising single slug bubble without a forced convection. In addition, void fraction was also measured in bubbly flow and bubbly-slug flow with a forced convection. Then, effects of flow patterns on the proposed estimations of void fraction and the accuracy of their estimations were discussed with the measurement results. From the result, the new estimations which consider a bubble shape are more accurate than the previous estimation in a slug bubble and bubbly-slug flow.     

Measurements of neutron induced activation of concrete at 64.5 MeV

The present paper presents the measurement of neutron induced activations on concrete using the 64.5 MeV quasimonoenergetic neutrons produced at the intense ⁷Li(p, n) neutron source at Cyclotron and Radioisotope Center, Tohoku Univeristy (CYRIC). The data were corrected for the effect of continuous neutrons in the source. The neutron energy, neutron yields and the spectrum of continuous neutrons were confirmed with the neutron time-of-flight method and the neutron activation measurement of the ²⁰⁹Bi(n, Xn) reactions having various threshold energy values. The nuclides produced by thermalized source neutrons are negligible. New data were obtained for concrete activation.     

快中子照相的实验研究

快中子照相是一种优良的无损检测技术,它有着X射线、热中子照相等所不具有的独特优势。本文介绍了作者分别采用^241Am-Be同位素中子源和中子发生器进行的快中子照相研究,并利用自制的Gd2O2S：Tb快中子转换发光屏进行探测,成功实现照相。     

On mechanisms by which a soft neutron spectrum may induce accelerated embrittlement

Both low displacement rates and softened neutron spectrum favor survival of a higher fraction of point defects per displacement for producing micro-structural changes leading to hardening and embrittlement. Low displacement rate results in low bulk recombination rate. A high thermal to fast neutron flux ratio results in a large fraction of point defects produced in small cascades from (n,γ) and (n,) reactions. Defects from such cascades generally avoid in-cascade recombination, while most of the defects created in large cascades produced by fast neutrons are lost to in-cascade recombination. Thus thermal neutrons produce more available defects per unit displacement dose. It is argued that the spectral effect may dominate the accelerated embrittlement observed in ferritic steels at the High Flux Isotope Reactor (HFIR) pressure vessel location. The rate effect is expected to be a secondary factor at temperatures as low as 50°C, where the HFIR data were obtained. Our analysis suggests generally that components subject to neutron environments with high thermal-to-fast ratios and irradiated at low temperatures may be subject to accelerated radiation effects.     

Development of an improved method for calculating the effective delayed neutron fraction of a core using a structure with only a few collapsed energy groups

In this study, an improved method for calculating the effective delayed neutron fraction of a core with a few collapsed energy groups has been developed. To accurately calculate the effective delayed neutron fraction of a core using the conventional method, a structure with several energy groups is needed for the fast energy region in order to reflect the difference in the fission spectra for prompt fission neutrons and delayed neutrons. On the other hand, a structure with only a few energy groups is needed for the criticality evaluation. Thus, the calculation cost increases for the effective delayed neutron fraction calculations owing to the need for a large number of energy groups. To solve this problem, in the present study, the error mechanism for the effective delayed neutron fraction calculation using a structure with only a few energy groups was studied, and it was found that the error results from the collapse of the fission spectra after the cell calculations without adjoint flux weighting. In addition, an improved method for the collapse fission spectra with an adjoint flux obtained by one-point calculation was developed. Using the proposed method, the effective delayed neutron fraction can be estimated with sufficient accuracy using a structure consisting of only a few collapsed energy groups. This result will contribute to reducing the calculation cost and/or improving the accuracy of effective delayed neutron fraction calculations.     

Mockup tests for developing the CARR-CNS with a two-phase thermo-siphon loop

The cold neutron source (CNS) is a facility to increase cold neutrons by scattering thermal neutrons in liquid hydrogen or deuterium around 20 K. For extracting a stable cold neutron flux from the CNS, the liquid quantity in the moderator cell should be maintained stably against disturbance of nuclear heating. The China Institute of Atomic Energy (CIAE) is now constructing the China Advanced Research Reactor (CARR: 60 MW). and designing the CARR-CNS with a two-phase thermo-siphon loop consisting of a condenser, two moderator transfer tubes and an annular cylindrical moderator cell. The mock-up tests were carried out using a full-scale loop with Freon-113,for validating the self-regulating characteristics of the loop, the void fraction less than 20% in the liquid hydrogen of the moderator cell, and the requirements for establishing the condition under which the inner shell has only vapor. The density ratio of liquid to vapor and the volumetric evaporation rate due to heat load are kept the same as those innormal operation of the CARR-CNS. The results show that the loop has the self-regulating characteristics and the inner shell contains only vapor, while the outer shell liquid. The local void fraction in the liquid increases with increasing of the loop pressure.     

可控中子源密度测井仪的密度响应特性与算法研究

在可控中子源密度测井中,脉冲中子源的中子产额存在波动,采用探测器绝对计数的密度算法受地层环境、测量条件的影响较大,因此会导致密度算法的精度较差。为提高密度测量的稳定性和精确性,本文针对可控中子源一体化测井仪的研发需要,建立了适合所研发仪器的密度算法：首先,结合蒙特卡罗数值模拟软件MCNP对新型一体化测井仪及地层进行了建模;然后,利用所建模型,模拟了在不同岩性、孔隙度条件下可控中子源发射的快中子与地层物质的相互作用过程,并通过记录γ探测器的近、远非弹性散射γ计数和俘获γ计数,中子探测器的近热中子及近超热中子计数信息,获得了一体化测井仪探测器计数与地层密度之间的响应关系,并对影响密度算法的主要因素进行了分析;最后,在密度响应特性分析的基础上,提出了使用近超热中子与近热中子计数比、近远非弹γ计数比来提高稳定性和精确性的可控中子源密度测井新算法。采用新密度算法对所建模型进行计算,结果表明,砂岩、灰岩、白云岩3种岩性下计算的密度与真密度非常接近,其相对误差小于6%。与哈里伯顿和斯伦贝谢算法的计算结果相比,本文方法显示了更好的效果：公式参数少、没有探测器的绝对计数、精确度高。     

Measurements of void fraction by an improved multi-channel conductance void meter

An improved multi-channel conductance void meter (CVM) was developed to measure a void fraction. Its measuring principle is based upon the differences in electrical conductance of a two-phase mixture due to the variation of void fraction around a sensor. The sensor is designed to be flush-mounted to the inner wall of the test section to avoid flow disturbances. The signal processor with three channels is specially designed so as to minimize inherent bias error due to the phase difference between channels. It is emphasized that the guard electrodes are electrically shielded in order not to affect the measurement of two-phase mixture conductance, but to ensure that the electrical fields are evenly distributed in the measuring volume. Void fraction is measured for bubbly and slug flow regimes in a vertical air–water flow and statistical signal processing techniques are applied to show that CVM has good dynamic resolution which is required to investigate the structural developments of bubbly flow and the propagation of void waves in a flow channel.     

Mass flow rate sensitivity and uncertainty analysis in natural circulation boiling water reactor core from Monte Carlo simulations

Our aim was to evaluate the sensitivity and uncertainty of mass flow rate in the core on the performance of natural circulation boiling water reactor (NCBWR). This analysis was carried out through Monte Carlo simulations of sizes up to 40,000, and the size, i.e., repetition of 25,000 was considered as valid for routine applications. A simplified boiling water reactor (SBWR) was used as an application example of Monte Carlo method. The numerical code to simulate the SBWR performance considers a one-dimensional thermo-hydraulics model along with non-equilibrium thermodynamics and non-homogeneous flow approximation, one-dimensional fuel rod heat transfer. The neutron processes were simulated with a point reactor kinetics model with six groups of delayed neutrons. The sensitivity was evaluated in terms of 99% confidence intervals of the mean to understand the range of mean values that may represent the entire statistical population of performance variables. The regression analysis with mass flow rate as the predictor variable showed statistically valid linear correlations for both neutron flux and fuel temperature and quadratic relationship for the void fraction. No statistically valid correlation was observed for the total heat flux as a function of the mass flow rate although heat flux at individual nodes was positively correlated with this variable. These correlations are useful for the study, analysis and design of any NCBWR. The uncertainties were propagated as follows: for 10% change in the mass flow rate in the core, the responses for neutron power, total heat flux, average fuel temperature and average void fraction changed by 8.74%, 7.77%, 2.74% and 0.58%, respectively.     

Effect of compensation error in drift-flux parameters on predictions of thermal-hydraulic parameters in nuclear safety system analysis codes

Void fraction in a nuclear reactor core is one of the most important parameters in a safety analysis using nuclear reactor thermal-hydraulics system analysis codes such as TRAC-BF1, RELAP5 and TRACE. Interfacial shear term governing void fraction in the two-fluid code is often estimated by Andersen approach which uses drift-flux type correlation to compute the interfacial shear term. The accuracy of two drift-flux parameters such as distribution parameter and drift velocity is anticipated to affect the accuracy of predicted void fraction significantly. In principle, the distribution parameter and drift velocity are independent parameters which should be determined by local gas and liquid velocities and void fraction. However, due to very limited local data, the distribution parameter and drift velocity are commonly determined by area-averaged void fraction and superficial gas and liquid velocities. This “approximate method” is acceptable when the distribution parameter and drift velocity are used together. However, in the Anderson's approach, the distribution parameter and drift velocity determined by the approximate method are used separately which may cause some compensation error in code calculations. In view of the great importance in accurately computing the interfacial shear term, the effect of the compensation error on the predicted void fraction is investigated. Intensive sensitivity analysis suggests the compensation error propagating to void fraction only up to 1% for steady state computations, whereas the effect of the compensation error on the predicted void fraction for transient computations becomes larger because temporal reduction of drag force may cause the increase in void fraction. A prototypic nuclear power plant analysis for ATWS scenario suggests that the overestimation of the void fraction may affect the neutron flux calculation.     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

基于CR39固体核径迹探测器的低能中子剂量测量方法初步研究

本文给出了一种采用固体核径迹探测器测量低能(0.5 eV以下能量)中子的双片径迹计数方法,利用蒙特卡罗模拟和实验,统计分析了低能中子引起的探测器单位面积径迹密度。结果显示,采用双片径迹计数方法,实现了CR39对于低能中子的响应,可以对低能中子引起的个人剂量进行估算。应用在中子职业个人剂量监测时,有效甄别了背景径迹的影响,因而降低了测量误差,对减小中子个人剂量监测的测量不确定度也是一个可行的方法。     

ST—451液体闪烁体中子探测器效率的测量

利用伴随粒子飞行时间方法，通过测量１４．７ＭｅＶ中子在聚乙烯和碳样品上的次级中子谱，得到了在０．９９－１４．７ＭｅＶ能量范围内ＳＴ－４５１液体闪烁体中子探测器的探测效率值，其总误差＜４．７％，其中统计误差＜３．５％。