压应力对各向同性细颗粒石墨热膨胀系数的影响

采用应变电测法测定了IG-11石墨在不同压应力下的热膨胀系数,研究了压应力对石墨热膨胀系数的影响.不同压缩载荷下的实验结果表明,压应力增大了平行于加载方向的热膨胀系数,并减小了垂直于加载方向的热膨胀系数.同时,随着载荷的不断增加,热膨胀系数的绝对变化量不断增大.此外,卸载后残余应变的存在也对热膨胀系数产生了类似的影响.     

基于氢研究核石墨中氚的去除

世界现阶段有大量的退役核石墨需要处理,~3H和~(14)C为其中含量最多、需重点去污核素。对于~(14)C来说,低温(不高于700°C)低氧环境下的热处理能比较有选择性地去除核石墨中的~(14)C。基于氚是氢的同位素、与氢具有相同的物理化学特性,本研究通过对三种不同产地的核石墨中氢在350oC的吸附以及400~700oC的解吸行为,探究核石墨中氚的去污工艺。实验发现:三种核石墨的氢吸附量不同,解吸规律大致相同,解吸量随时间的变化上有差异。国产核石墨NG-CT-10、日本核石墨IG-110以及德国核石墨NBG-18的氢总吸附量分别为6.7×10-3 mL·g-1、9.30×10~(-3) mL·g~(-1)以及9.12×10~(-30 mL·g~(-1),其中化学吸附量分别为3.2×10~(-3) mL·g~(-1)、3.0×10~(-3) mL·g~(-1)和0.92×10~(-3) mL·g~(-1)。石墨对氢吸附量上的差异可能来源于三种核石墨的不同制备工艺和物理性质上的一些差异,这些差异主要来自于平均孔径、比表面积、成型工艺以及焦粒粒径上的区别;NG-CT-10有效吸附量所占比最高,表明NG-CT-10有较大量的氚吸附量。400~700oC的核石墨氢解吸实验表明:三种石墨中的氢主要是从700oC开始有效解吸,但各自相对于总吸附量的解吸量有明显区别,NG-CT-10、IG-110和NBG-18在700°C时的解吸量分别为7%、13.5%和70%。由此可得,NBG-18中的氚最易被解吸出来。根据氢在石墨中的吸附模型,700oC解吸出来的氢应该位于石墨晶粒边缘。为了解吸剩余氚,同时不影响~(14)C的有效去除,不提高热处理温度,可能需要改变解吸时的载气组分。     

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。     

Behaviors of fine (IG-110) and ultra-fine (HPG-510) grain graphite irradiated by 7 MeV Xe26+ ions

Developing a molten salt reactor needs molten salt-impermeable nuclear graphite. Ultra-fine grain graphite is a good choice as it is better in permeability than fine grain graphite. In this paper, ultra-fine grain graphite(HPG-510) and fine grain graphite(IG-110) samples are irradiated at room temperature by 7 MeV Xe ions to doses of 1 × 10~(14)-5 × 10~(15) ions/cm~2. Scanning electron microscopy, transmission electron microscopy(TEM), Raman spectroscopy and nano-indentation are used to study the radiation-induced changes. After irradiation of different doses, all the HPG-510 samples show less surface fragment than the IG-110 samples. The TEM and Raman spectra,and the hardness and modulus characterized by nano-indentation, also indicate that HPG-510 is more resistant to irradiation.     

化学光谱法测定二氧化镎中八个杂质元素

本文介绍了测定二氧化镎中八个杂质元素(铁、锰、镁、镍、铬、铜、钠和钾)的化学光谱法。样品经溶解、还原后采用硝酸型阴离子交换树脂除去溶液中99.9 %以上的镎,以交流电弧溶液干渣法光谱测定流出液中的杂质元素。测定范围:铁、镁5×10~(-5)%—1.2×10~(-3)%;镍、铬2.5×10~(-5)%—6×10~(-4)%;锰、铜2.5×10~(-6)%—6×10~(-5)%;钠、钾1×10~(-4)%—2.4×10~(-3)%。杂质元素的回收率为82—110%,方法的精密度小于20%。     

基于同步辐射原位拉伸XRD研究熔盐浸渗的核石墨IG-110微观结构演化

钍基熔盐堆(Thorium Molten Salt Reactor,TMSR)是第四代核反应堆的代表之一,其特点是以熔融氟盐作为冷却剂和燃料的载体。在熔盐堆中,熔盐容易浸渗到核石墨内部,引发核石墨局部高温,造成核石墨损伤程度增加,严重破坏核石墨的结构,从而影响核石墨材料的宏观性能和使用寿命。然而,熔盐浸渗对核石墨力学性能的微观机制以及熔盐浸渗引起的微结构损伤或破坏机制目前仍不清晰,因此有待进一步研究原位环境下(如力学加载、高温等)熔盐浸渗对核石墨微结构的影响,并揭示微结构演化的相关机制。本文基于同步辐射原位拉伸X射线衍射技术(Two Dimensional X-ray Diffraction,2D-XRD),开展了外部载荷为0 N、15 N、21 N、27 N和32 N时熔盐浸渗后的核石墨IG-110在拉伸断裂过程中的微观结构演化研究,以揭示外部载荷条件下的核石墨IG-110与熔盐之间的原位实时相互作用及材料断裂的微观机制。实验结果表明:在拉伸断裂过程中外部载荷使熔盐浸渗后的核石墨IG-110的结晶性变差、层间距变大,同时FLiNaK盐的结晶性也明显变差。这一发现将有助于解释熔盐浸渗后核石墨IG-110力学性能的变化,理解核石墨IG-110与FLiNaK熔盐间的相互作用机理,有利于高性能核石墨的制备和TMSR的安全可靠运行分析。     

石墨的热中子吸收截面测量

本文利用脉冲中子源法测量了一批石墨的热中子吸收截面。测量的几何曲率范围是(0.699—13.26)×10~(-3)厘米~(-2)。由于几何曲率在6.341×10~(-3)厘米~(-3)以上时,实验上得不到恒定的衰减常数,所以用来求热中子吸收截面的几何曲率范围为(0.699—6.341)×10~(-3)厘米~(-2)。石墨密度为1.653克/厘米~3,测量温度为14.5℃。求得石墨热中子吸收截面σ_α=4.03±0.13毫靶,扩散系数D_0=(2.073±0.025)×10~5厘米~2·秒~(-1),扩散冷却系数C=(3.17±0.49)×10~6厘米~4·秒~(-1)。换算到标准条件(密度1.60克/厘米~3,温度20℃),D_0=(2.160±0.026)×10~5厘米~2·秒~(-1),C=(3.41±0.51)×10~6厘米~4·秒~(-1)。     

CRDM钩爪用stellite-6合金冲击磨损性能研究

采用自制的试验机模拟核电厂一回路水质,研究不同温度和应力下控制棒驱动机构(CRDM)钩爪用stellite-6合金的冲击磨损性能。研究结果表明:正常接触条件下,温度对冲击磨损性能的影响较小,合金尺寸变化速率约为1×10~(-8)mm/次,质量变化速率约为2×10~(-6) mg/次;高接触应力下,合金的冲击磨损机制为塑性变形和疲劳剥落;室温、90℃和150℃下,合金尺寸变化速率分别为:1.3×10~(-7) mm/次、4.7×10~(-7) mm/次和5.3×10~(-7) mm/次,质量变化速率分别为7.5×10~(-6) mg/次、4.17×10~(-5) mg/次和4.83×10~(-5) mg/次。     

核级石墨失重率对其氧化速率的影响

采用热重分析方法研究600⁷50℃温度范围内IG-110核级石墨在空气中的氧化行为,分析失重率对其氧化速率的影响。利用随机孔隙模型拟合了IG-110、H-451、NBG-18及NBG-10等4种核级石墨的结构参数,并确定了各自的失重率影响因子表达式。结果表明,随着氧化反应的进行,核级石墨的氧化速率随失重率的增大呈现先增大后减小的趋势,氧化速率的最大值出现在30%750℃温度范围内IG-110核级石墨在空气中的氧化行为,分析失重率对其氧化速率的影响。利用随机孔隙模型拟合了IG-110、H-451、NBG-18及NBG-10等4种核级石墨的结构参数,并确定了各自的失重率影响因子表达式。结果表明,随着氧化反应的进行,核级石墨的氧化速率随失重率的增大呈现先增大后减小的趋势,氧化速率的最大值出现在30%⁴0%失重率范围。使用随机孔隙率模型可以较好地模拟失重率对氧化速率的影响,其中石墨结构参数随核级石墨平均粒径的增加而减少。     

裂变径迹年龄测定中存在的一些问题

裂变径迹年龄测定方法在地质学、考古学中已经广泛地得到了应用。本文着重讨论该方法目前存在的一些问题和今后发展的方向。 目前存在的问题主要有:1.~(238)U自发裂变的衰变常数值。目前已发表的~(238)U自发裂变衰变常数(λ_f)值约40个以上,在裂变径迹年龄测定中采用的λ_f值有两组:6.9×10~(-7)年~(-1)(或6.85×10~(-17)年~(-1),7.00×10~(-17)年~(-1))和8.4×10~(-17)年~(-1)。虽然大多数实验室采用前者,但仍有少部分实验室坚持采用后者。2.样品在     

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.     

The effect of stress levels on the coefficient of thermal expansion of a fine-grained isotropic nuclear graphite

Due to the fluctuation and non-uniform distribution of temperature within the core structure of high-temperature gas-cooled reactors (HTGRs), the thermal expansion behavior of graphite materials plays an important role in the design of graphite components, especially of large-scale components. In the present paper, in order to investigate the influence of stress levels on the coefficient of thermal expansion (CTE) of IG-110 graphite, the strain gauge method was used to measure the CTE on the cylindrical specimens under a series of loads applied using a universal tensile testing machine. In addition, a more precise measurement using a thermal dilatometer was employed to validate the tests using the strain gauge method. A good agreement has been obtained between the experimental results using these two methods. The results show that when the specimens were under compressive loads, the CTE along the loading direction of the specimens increased and that along the perpendicular direction decreased, with more changes in the former. The absolute changes of the CTE in the two directions increased with increasing applied load. When graphite specimens were subjected to a compressive load of 40 MPa, the axial CTE of specimens sectioned along the radial direction of the graphite brick as it is installed in the core structure increased from 4.13 × 10⁻⁶ to 5.35 × 10⁻⁶ K⁻¹, while the axial CTE of specimens sectioned along the vertical direction increased from 3.97 × 10⁻⁶ to 5.58 × 10⁻⁶ K⁻¹. Moreover, the residual change of the CTE, which was caused by the permanent residual strain after unloading, was observed. The change of the CTE with stress levels should be considered in the stress analysis and life prediction of the nuclear graphite components.     

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.     

CO2 corrosion of IG-110 nuclear graphite studied by gas chromatography

The CO₂ corrosion behavior of IG-110 nuclear graphite has been investigated using the gas chromatography method which allows the continuous analysis of the CO₂/CO gas mixture at the outlet of the corrosion chamber. The effects of temperature and initial CO₂ concentration are studied based on the Arrhenius-type reaction model. From 745 to 995 °C, the Arrhenius curve shows a linear behavior. For higher temperatures, a non-linear behavior is observed. The activation energy is calculated as 210 kJ/mole and is independent of the initial CO₂ inlet concentrations of 10%, 14% and 17%. The corrosion behavior at 1145 °C, in the diffusion-controlled regime, has also been investigated. At this temperature, the interior of IG-110 graphite is severely attacked by CO₂, and the material's surface morphology is changed drastically. A measurement of the corrosion rate against corrosion time shows that the corrosion rate initially increases to a maximum value at a weight loss degree of 30%–35%, after which it begins to decline.     

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.     

Precision dilatometer analysis of neutron-irradiated nuclear graphite recovery process up to 1673 K

Specimens of two kinds of isotropic nuclear graphite, IG-110U and ETP-10, were neutron-irradiated at fluence of 1.92 × 10²⁴ n/m² (E > 1.0 MeV) at 473 K. The recoveries of the macroscopic lengths of these specimens during isothermal and isochronal annealing at temperatures of up to 1673 K were investigated in a step-wise manner by using a precision dilatometer. The macroscopic lengths after isochronal annealing for 6 h at each temperature decreased gradually as the temperature was increased to 1673 K. The recovery trends of the c-axis and a-axis lattice parameters differed from one another, and from the macroscopic length recovery trends. For the IG-110U specimen, the activation energies (E_a) of macroscopic volume recovery corresponding to annealing at 523–773, 773–923, 923–1073, and 1073–1173 K were found to be 0.15, 0.34, 0.73, and 2.59 eV, respectively. For the ETP-10 specimen, the E_a corresponding to 523–923, 923–1223, and 1223–1373 K were determined to be 0.15, 0.46, and 2.19 eV, respectively. These results indicate that both graphite specimens underwent three or four stages of macroscopic length recovery between 523 K and the annealing temperatures at which their initial lengths were recovered. It is suggested that during the first stage recovery proceeded via the migration of single interstitials along the basal plane and the resulting V-I recombination. In the middle stages, recovery occurred due to the migration of interstitial groups such as C₂ along the basal plane, while in the last stage, it proceeded via through-layer migration of interstitials or migration of single vacancies.     

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.     

Investigation of kinetic recovery process in low dose neutron-irradiated nuclear graphite by thermal annealing

To investigate the kinetic recovery process of low dose neutron-irradiated graphite, nuclear-grade isotropic graphite IG-110U and ETP-10 were neutron irradiated using the JMTR up to 1.38 × 10²³ n/m² (E_n > 1 MeV) at ~473 K. In-situ measurement of macroscopic length was conducted during the isothermal and isochronal annealing process from room temperature up to 1673 K. From room temperature to 773 K for IG-110U, and to 1023 K for ETP-10, macroscopic lengths, lattice parameters, and unit cell volumes of both specimens recovered to their pre-irradiation values, and this recovery process subdivided into two stages. The activation energies of macroscopic volume recovery at 523–673 K and 673–773 K were determined to be ~0.22 eV and ~0.57 eV for IG-110U, respectively; ~0.13 eV and ~2.59 eV at 523–923 K and 923–1023 K for ETP-10, respectively. The migration of not only single interstitials but also interstitials dissociated from submicroscopic interstitial groups along basal planes followed by vacancy-interstitial recombination play a dominant role in the first stage. The second stage is suggested to proceed via the motion of carbon groups along basal planes for IG-110U, and migration of single interstitials along the c-axis for ETP-10. During 773 K or 1023 K up to 1673 K, macroscopic length continuously shrank with decreasing shrinking rate, even with a turnaround to swell at 1173 K for IG-110U.     

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.