Oxidation behavior of IG and NBG nuclear graphites

This work studies the oxidation-induced characteristics of four nuclear graphites (NBG-17, NBG-25, IG-110, and IG-430). The oxidation characteristics of the nuclear graphites were measured at 600 °C. The surface properties of the oxidation graphites were characterized by means of scanning electron microscopy, X-ray photoelectron spectroscopy, and contact angle methods. The N₂/77 K adsorption isotherm characteristics, including the specific surface area and micropore volume, were investigated by means of BET and t-plot methods. The experimental results show an increase in the average pore size of graphites; they also show that oxidation produces the surface functional groups on the graphite surfaces. The surface area of each graphite behaves in a unique manner. For example the surface area of NBG-17 increases slightly whereas the surface area of IG-110 increases significantly. This result confirms that the original surface state of each graphite is unique.     

Photon interaction with semiconductor and scintillation detectors

Molten salt is used as primary coolant flowing through graphite moderator channel of a molten salt reactor.Working at high temperature under radiation environment,the pore network structure of nuclear graphite should be well understood.In this paper,X-ray tomography is employed to study the 3D pore structure characteristics of nuclear grades graphite of IG-110,NBG-18 and NG-CT-10,and permeability simulation through geometries are performed.The porosity,number of pores and throats,coordination number and pore surface are obtained.NGCT-10 is of similar microstructure to IG-110,but differs significantly from NBG-18.The absolute permeabilities of IG-110,NG-CT-10 and NBG-18 are 0.064,0.090 and0.106 mD,respectively.This study provides basis for future research on graphite infiltration experiment.     

ASME规范概率评价方法对细颗粒石墨的适用性研究

石墨由于其高中子散射截面和低中子吸收截面特性,被广泛应用于第四代高温气冷堆中作为慢化剂、反射层和堆芯结构,故保证其结构完整性对反应堆的安全运行非常重要。由于石墨材料强度分散,概率论方法评价其失效较常用的确定论评价方法更为合适。目前,美国ASME规范采用的概率方法主要针对NBG-18这种大颗粒石墨,对我国高温气冷堆核电站工程项目采用的细颗粒石墨IG-110的适用性未知。同时,我国成都碳素生产的高温堆备选石墨NG-CT-01颗粒大小与IG-110相似,也为细颗粒石墨。因此,文章研究ASME规范概率方法对细颗粒石墨的适用性,并通过实验数据加以验证。结果表明,对于细颗粒石墨,ASME规范过于保守,低估了材料的强度性能。     

Effects of reaction temperature and inlet oxidizing gas flow rate on IG-110 graphite oxidation used in HTR-PM

The oxidation behavior of a selected nuclear graphite (IG-110) used in Pebble-bed Module High Temperature gas-cooled Reactor was investigated under the condition of air ingress accident. The oblate rectangular specimen was oxidized by oxidant gas with oxygen mole fraction of 20% and flow rates of 125–500 ml/min at temperature of 400–1200?°C. Experiment results indicate that the oxidation behavior can also be classified into three regimes according to temperature. The regime I at 400–550?°C has lower apparent activation energies of 75.57–138.59 kJ/mol when the gas flow rate is 125–500 ml/min. In the regime II at 600–900?°C, the oxidation rate restricted by the oxygen supply to graphite is almost stable with the increase of temperature. In the regime III above 900?°C, the oxidation rate increases obviously with the increase of temperature. With the increase of inlet gas flow from 125 to 500 ml/min, the apparent activation energy in regime I is increased and the stableness of oxidation rate in regime II is reduced.     

Anisotropy of coefficient of thermal expansion of nuclear graphite under compressive stresses

The coefficient of thermal expansion (CTE) of nuclear graphite IG-110 and NBG-18 under compressive stresses of 20 MPa, 30 MPa and 40 MPa has been measured by strain gauge method and corresponding anisotropies of CTE under stresses were investigated. With the increasing compressive stresses, the CTE of IG-110 and NBG-18 parallel and perpendicular to the loading directions increased significantly and decreased gradually respectively. The corresponding CTE anisotropies of IG-110 and NBG-18 almost maintain below 1.05 and keep their original near-isotropic properties under compressive stresses maybe due to the homogeneous sensitivity of CTE to the stresses, perfect crystallites in the grains and homogeneous alignment of grains in graphite. The constant isotropic properties of graphite IG-110 and NBG-18 under stresses are beneficial for the integrity and safety of the graphite used in the reactor.     

Improved Graphite Damage Model for Predicting Property Changes of HTGR Graphites under Isothermal and Nonisothermal Irradiations

Graphite components of a High Temperature Gas-Cooled Reactor (HTGR) core will be subjected to fast neutron irradiation-induced damages causing changes in engineering properties which are dependent significantly on irradiation temperature and neutron fluence. A Graphite Damage Model (GDM) elaborated to interpret observed isothermal irradiation behavior of various polycrystalline graphites has been improved in order to apply it to nonisothermal irradiations. The paper outlines the physical and mathematical formulations of the improved GDM together with some comparisons between predictions and measurements. On the basis of nonlinear fittings of the model functions to measured data, a total of 28 global modeling parameters have been determined successfully. Furthermore, a recurrence formula has been devised to permit a nonisothermal irradiation behavior to be predicted in terms of the new GDM. This has been proved by comparison of the predicted changes in linear dimension and thermal conductivity with the measured ones of some graphite materials irradiated at constant and changing temperatures.     

CFD simulation of particle deposition on an array of spheres using an Euler/Lagrange approach

The Generation IV Pebble Bed Modular Reactor (PBMR) is being considered as a promising concept to produce electricity or process heat with high efficiencies and unique safety features. The PBMR is a high-temperature, helium-cooled, graphite moderated reactor. The fuel elements consist of 6 cm diameter spherical graphite “pebbles” containing each thousands of uranium dioxide microspheres.As the pebbles continually rub against one another in the core, a significant quantity of graphite dust can be released in the reactor coolant system. These dust particles, which contain some amounts of fission products, are transported and deposited on pebbles as well as primary circuit surfaces. It is therefore of great safety interest to develop and benchmark numerical approaches for predicting deposition of dust particles in the various locations of the PBMR primary circuit.In this investigation, we use the ANSYS-Fluent CFD code to simulate particulate flows around linear arrays of spheres and compare deposition rates against experiments. It is found that the Reynolds Stress Model (RSM) combined with the Continuous Random Walk (CRW) to supply fluctuating velocity components predicts deposition rates that are generally within the scatter of the data. The methodology developed here can therefore be used to predict to first order the graphite dust deposition rates on pebbles in PBMR-type reactors.     

Kinetics and mechanisms of the reaction of air with nuclear grade graphites: IG-110

The work presented in this report is part of an ongoing effort in the microgravimetric evaluation of the intrinsic reaction parameters for air reactions with graphite over the temperature range of 450 to 750°C. Earlier work in this laboratory addressed the oxidation/etching of H-451 graphite by oxygen and steam. This report addresses the air oxidation of the Japanese formulated material, IG-110. Fractal analysis showed that each cylinder was remarkably smooth, with an average value, D, the fractal dimension of 0.895. The activation energy, E_a, was determined to be 187.89 kJ/mol indicative of reactions occurring in the zone II kinetic regime and as a result of the porous nature of the cylinders. IG-110 is a microporous solid. The low initial reaction rate of 9.8×10⁻⁵ at 0% burn-off and the high value (764.9) of Φ, the structural parameter confirm this. The maximum rate, 1.35×10⁻³ g/m²s, was measured at 34% burn-off. Reactions appeared to proceed in three stages and transition between them was smooth over the temperature range investigated. Both E_a and ln A did not vary with burn-off. The value of ΔS, the entropy of activation, was −41.4 eu, suggesting oxygen adsorption through an immobile transition state complex. Additional work is recommended to validate the predictions that will be made in relation to accident scenarios for reactors such as the modular high temperature gas-cooled reactor where fine grained graphites such as IG-110 could be used in structural applications.     

Model of particle resuspension in turbulent flows

The graphite dust generated in an HTR/PBMR during normal reactor operation is deposited inside the primary system and becomes radioactive due to sorption of fission products. A significant amount of radioactive dust may be resuspended and released to the environment in case of LOCA. Therefore accurate particle resuspension models are required for HTR/PBMR safety analyses.Thermal-hydraulic safety analyses of HTR/PBMR type reactors are typically performed using computer codes such as FLOWNEX, MELCOR, or SPECTRA. None of these codes currently includes a well-tested mechanistic resuspension model.The resuspension model based on the Vainshtein model has been developed and implemented into the SPECTRA thermal-hydraulic system code. The resuspension model formulation has been extended in such way that other formulations, for example the Rock’n Roll model, may easily be defined and used within the general model framework.Several test calculations were performed, including comparisons of the numerical SPECTRA results with the analytical solutions obtained by means of MathCAD. Furthermore, comparisons with the experimental results of the Reeks and Hall, and STORM experiments were made. It was concluded that the model gives satisfactory results for a number of tests.     

基于SAXS方法的核石墨微孔结构的温度影响

X射线小角散射(Small Angle X-ray Scattering,SAXS)是研究纳米尺度微观结构的重要手段。本文利用同步辐射SAXS技术测量了25oC、100oC、200oC、300oC和400oC时,IG-110和NBG-18核石墨在纳米尺度范围内孔隙的数量分布及其分形特征的变化。实验结果表明,IG-110和NBG-18核石墨的微观结构中存在微小尺寸上的不均匀区域,且核石墨孔隙的固气结构具有明锐的界面。但随着温度的升高,固气界面的变化并没有呈现出明显的规律性。此外,在纳米尺度上,IG-110和NBG-18核石墨的孔隙数量随温度呈现增加的趋势,且IG-110核石墨孔隙数量的增加幅度大于NBG-18核石墨,其平均孔隙尺寸的减小幅度大于NBG-18核石墨。在核石墨的微孔结构内,其固气界面的分形维数随温度升高逐渐减小,且NBG-18核石墨分形维数的变化幅度小于IG-110核石墨。这表明核石墨的分形结构随温度的升高逐渐光滑。     

Comparison of oxidation model predictions with gasification data of IG-110, IG-430 and NBG-25 nuclear graphite

A phenomenological oxidation kinetics model of graphite is presented and its results are compared with the reported experimental gasification data for nuclear graphite of IG-110, IG-430 and NBG-25. The model uses four elementary chemical kinetics reactions, employs Gaussian-like distributions of the specific activation energies for adsorption of oxygen and desorption of CO gas, and accounts for the changes in the effective surface areas of free active sites and stable oxide complexes with weight loss. The distributions of the specific activation energies for adsorption and desorption, the values of the pre-exponential rate coefficients for the four elementary chemical reactions and the surface area of free active sites are obtained from the reported measurements using a multi-parameter optimization algorithm. At high temperatures, when gasification is diffusion limited, the model calculates the diffusion velocity of oxygen in the boundary layer using a semi-empirical correlation developed for air flows at Reynolds numbers ranging from 0.001 to 100. The model also accounts for the changes in the external surface area, the oxygen pressure in the bulk gas mixture and the effective diffusion coefficient in the boundary layer with weight loss. The model results of the total gasification rate and weight loss with time in the experiments agree well with the reported measurements for the three types of nuclear graphite investigated, at temperatures from 723 to 1226 K and weight loss fractions up to ～0.86.     

Interpolation and Extrapolation Method to Analyze Irradiation-Induced Dimensional Change Data of Graphite for Design of Core Components in Very High Temperature Reactor (VHTR)

Graphite materials are used as core components in the High-Temperature Gas-Cooled Reactor (HTGR) and Very High Temperature Reactor (VHTR). The authors prepared technical documents for design, material, products, in-service inspection and maintenance of the graphite components for the HTGR/VHTR, which were summarized as a draft of standard for the graphite components through discussion made in a “Special committee on research on preparation for codes for graphite components in HTGR” set up within AESJ. The draft of standard contains graphical expressions for the irradiated material properties of IG-110 graphite. It is possible to use the graphical expressions for the components design of VHTR. The graphs were obtained based on the interpolation and extrapolation of the irradiation data. The irradiation-induced dimensional change of IG-110 graphite was obtained through the interpolation and extrapolation of the irradiation data with a quadratic equation of fast neutron fluence. The irradiation data for H-451 and ATR-2E graphites were used for the evaluation of the interpolation and extrapolation of irradiation data for IG-110. It was shown in this study that the proposed interpolation and extrapolation method is reasonable for IG-110 with regard to the database available at present.     

掺杂石墨在高能激光束和电子束作用下的热冲击行为

石墨被广泛用于当今的托卡马克装置中 ,作为真空室第一壁和偏滤器靶板的保护材料 ,也是未来聚变堆的一种候选面对等离子体材料。其抗化学溅射性能和抗热冲击性能受到广泛关注。用高能激光束和电子束轰击实验材料 ,模拟聚变堆面对等离子体材料在等离子体破裂时的工作状态 ,考察了 4种掺杂石墨材料在热冲击下的热腐蚀规律。实验结果表明 ,石墨掺杂能有效降低材料的烧蚀率。当激光单脉冲能量密度为 491 5KJ m2 时 ,冲击频率 1 0Hz,持续辐照 3 0秒后 ,几种掺杂石墨的失重率不超过2 1 3 6mg cm2 ,表现出了比纯石墨更优良的抗热冲击性能。     

Change in electrical resistivity of nuclear graphite during compression tests and a model for its deformation and fracture mechanism

The results of the resistivity changes during compression of some nuclear graphites are summarized in order to cast light on the fracture mechanism of the materials; data on pyrolytic graphite and amorphous carbon are also taken into account. It is found that all the graphites investigated show an abrupt increase in resistivity when the applied stress increases to about a half of the fracture stress. Above this stress the non-linearity of the stress-strain curve becomes more pronounced and the formation and growth of optically resolvable cracks occur. A model based on the deformation of cracks and pores on the basal plane is proposed for explaining the change in resistivity, and is supported by measurements of the effect of pre-stressing on the Young's modulus, thermal expansion, mercury porosimetry and Knoop micro-hardness of the material.     

Adsorption of iodine on graphite in High Temperature Gas-Cooled Reactor systems: A review

Adsorption of iodine on graphite is of great interest for operation and safety of high temperature nuclear reactors. Graphite can adsorb significant amounts of iodine and retain it for a long period of time. Significant amount of work on this subject has been done in the past. Various types of adsorption apparatus have been designed and data were collected. The types of graphite used in past studies are not available anymore, and as a consequence the data are not applicable for the new type of commercial nuclear grade graphites. However, the past experimental systems, data, and their analysis are useful to design a better experimental system, collect more accurate data, and, provide better understanding of the adsorption process and data. In addition the existing data can be used to generate a framework to understand the types of adsorption processes taking place. In this work, we have conducted an exhaustive literature review and further analyzed the data. Four adsorption isotherms; the Langmuir, the Freundlich, and the two isotherms proposed in the International Atomic Energy Agency (IAEA) Tecdoc-978 were used to correlate the available equilibrium adsorption data. For most of the data, the simple Langmuir and the Freundlich isotherms provided a reasonable fit of the data. The Polyani's potential theory was also used to check the consistency of the data and as indicated by the theory, most of the data set provided a single characteristic curve. The isosteric heats of adsorption calculated using the literature data suggested that iodine-adsorption on graphite could be a chemisorption process.     

底反射层氧化对模块式高温气冷堆进气事故的影响分析

进气事故是模块式高温气冷堆(HTR-PM)事故分析中重点考虑的一种事故类型。核级石墨在高温气冷堆中被广泛用作反射层材料、结构材料和慢化材料等。在进气事故中,燃料元件基体石墨发生氧化反应增加了燃料颗粒裸露和放射性释放的风险,底反射层发生氧化反应降低了石墨材料的机械性能,可能破坏堆芯底部结构的完整性。本文利用高温气冷堆专用系统分析程序TINTE,分别选取两种不同氧化速率的石墨材料作为底反射层材料,以热气导管双端断裂的进气事故为例,分析不同材料对进气事故的影响。在保证底反射层完整性的前提下,底反射层采用高氧化速率的材料时,能明显降低燃料颗粒裸露和放射性释放的风险。     

Oxidative erosion of graphite in air between 600 and 1000 K

The oxidative erosion of seven types of graphite has been investigated by heating in air at temperatures between 600 and 1000 K. The specimens include pyrolytic graphite, fine-grain graphites, carbon-fibre composites (CFC), and graphites doped with Si and Ti. The weight loss was measured using a microbalance, the surface morphology by scanning electron microscopy, and the composition of the surface layer by MeV ion beam techniques. Pyrolytic graphite is least affected by erosion, while pure and Si-doped CFCs erode particularly fast. Typical erosion rates for specimens with a surface area of ?4 cm² are below 0.2 μg/m² s at 600 K for all graphite types, and at 900 K range from 0.34 mg/m² s for pyrolytic graphite to about 9 mg/m² s for the strongest eroding types. The temperature dependence of the erosion rate of all types of graphite studied is well described by an activation energy of 1.7 eV. The erosion rates of these graphites are by far lower than the removal rates for deposited amorphous hydrocarbon layers. In contrast to all other types, the Ti-doped graphite absorbs a significant amount of oxygen reaching up to ?5% of its original mass. Once the oxygen uptake is saturated, it erodes with rates similar to those of the strongest eroding types.     

Investigation of oxidation behaviors of nuclear graphite being developed and IG-110 based on gas analysis

The oxidation behaviors of the nuclear graphite being developed were investigated using gas chromatograph at 873–1373 K. The oxidation experiments were carried out with the gas flow rate of 0.2 L/min and the oxygen concentrations of 7, 10 and 20 mol%. The oxidation reaction began at 973 K and was accelerated with the increase of temperature. At 1173–1273 K, the oxidation was limited by oxygen supplied to graphite and the reaction rate held steady. From 1273 to 1373 K, the oxidation rate increased obviously due to the significant reaction between CO₂ and graphite. At the low temperature regime (973–1073 K), the apparent activation energies with the oxygen mole fractions of 7%, 10% and 20% were 298, 324 and 321 kJ/mol, respectively. Scanning electron microscope was applied to reveal the pore development of the graphite oxidized at different temperatures. The effect of CO combustion at temperature below 1173 K was discussed based on the oxidation behaviors of the graphite being developed and IG-110. It was suggested that the ASTM D7542-15 standard should be adjusted to fit some popular graphite, such as graphite IG-110.     

Penetration depth and transient oxidation of graphite by oxygen and water vapor

Equations are derived for the approach to equilibrium in the oxidation of graphite under assumptions of constant graphite density and linearized oxidation kinetics. A two-factor expression is assumed for the effective diffusivity. Equilibration may be estimated by observing the convergence of profiles with time or by means of an algebraic approximation. At large times, the profiles converge to the steady state. Oxidation depths show fair agreement with published measurements and follow closely the observed temperature trend.     

Analysis of As-Machined Nuclear Graphite Parts Using Conventional and Laser Ultrasonic Techniques

The feasibility of remote, in-service monitoring of as-machined nuclear graphite parts using noncontacting laser ultrasonics was investigated. The anisotropy sensitivities of laser ultrasonics and several other conventional techniques were compared for three significantly different nuclear- grade graphites. Laser ultrasonics using a Fabry-Perot interferometer proved sufficiently sensitive to detect the anisotropy in a bulk graphite having moderate grain size. In contrast, a coarsegrained graphite was difficult or impossible to investigate using any of the techniques. Laser ultrasonic monitoring of some graphites may be useful in future nuclear operations, particularly for in-service applications where contacting and destructive analyses are unsuitable.