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。     

HTR-10中石墨材料氧化的估算

通过合理的分析模型,研究了10MW高温气冷堆（HTR-10）中石墨材料IG-11的氧化情况。以HTR-10反射层为例,计算了反射层最高温度处氧气对石墨的氧化情况。计算发现,石墨的氧化程度随反应产物中CO份额的增加而加剧;当反应全部为CO时,石墨材料的最大理论耗蚀厚度为35．8mm;当反应全部为CO2时,石墨材料的最小理论耗蚀厚度约为23．7mm;石墨材料在整个服役期间,在反射层温度最高处,石墨的氧化比较严重。     

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.     

HTR-10一回路放射性核素在石墨上的吸附及其微观机理研究

通过巨正则系综方法与第一性原理计算,研究了100～900 ℃下,¹³⁴Cs、¹³⁷Cs、⁹⁰Sr、¹¹⁰Ag^m、¹³¹I 5种重要核素在石墨上的吸附率随温度、压强等参数的变化,并根据工程实际参数,推算他们在HTR-10一回路中反射层、石墨碳砖以及石墨粉尘上的吸附量。研究表明,Cs和Sr倾向于吸附在石墨的H位,而Ag和I倾向于吸附在石墨的T位,且他们的吸附能也有所差异。此外,核素粒子数密度与吸附率呈线性关系,而温度与吸附率呈指数关系。最后,通过研究5种核素在HTR-10一回路中的吸附情况,发现其中的放射性主要来自于核素¹³⁴Cs、¹³⁷Cs和¹³¹I,而⁹⁰Sr和¹¹⁰Ag^m的贡献较少,这与唯象模型的保守估计结论一致。     

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.     

HTR-10辐照后石墨球~(14)C含量研究

高温气冷堆中球形燃料元件与不含燃料的石墨球在14 C产生机制上基本相同,为获得10MW高温气冷堆(HTR-10)中燃料元件及石墨球中14 C的含量,研究了14 C在石墨球中的产生机理。总结了石墨球及燃料元件中14 C的产生途径,计算经过堆芯辐照后的石墨球中14 C总量,比较现有石墨球的解体技术,提出了分解石墨球制取14 C样品的实验测量方法。本文工作为进一步的实验研究工作奠定基础并提供理论计算结果比对。     

HTR-10 GT辅助轴承保持架振动特性研究

在10 MW高温气冷堆氦气汽轮机发电系统(HTR-10 GT)中,辅助轴承作为磁力轴承支承失效后转子的辅助支承装置,是整个转子系统最重要的安全保障.针对辅助轴承的工作特点,采用有限元方法进行了建模分析,研究了辅助轴承中保持架的离心应力和自由振动特性,讨论了不同结构参数对保持架振动特性的影响.结果表明,保持架上的最大离心应力发生在侧梁中点处;在辅助轴承工作状态下,较易引发低频的振动模态;对保持架尺寸以及兜孔数的合理选择有助于提高保持架的性能.     

High temperature reactor development in China

Since the late 1970'-s the research and development program on the high temperature gas-cooled reactor (HTR) has been carried out in China. The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) reached first criticality in 2000 and was put into full power operation in 2003. Six safety demonstration tests were done on the HTR-10. The project of the HTR-10 with a gas turbine cycle is underway. The project of the HTR demonstration plant with a power of around 150 MWe (HTR-PM) is planned. In this paper the HTR development in China is briefly described.     

HTR-10放射性石墨粉尘取样回路中过滤元件的效率研究

10MW高温气冷实验堆(HTR-10)中石墨构件摩擦磨损不可避免地会产生石墨粉尘,其上可吸附固体裂变核素,可能影响反应堆安全正常运行。为研究石墨粉尘的粒径分布、微观形貌以及其表面的吸附特性,在HTR-10一回路氦净化系统入口建造了配置高性能烧结金属粉末过滤元件的放射性石墨粉尘取样回路。为对放射性石墨粉尘的取样过滤结果进行研究,首先对过滤元件的过滤机理、压降特征以及过滤效率进行分析。设计并搭建实验装置对烧结金属粉末过滤元件的过滤效率进行测量,实验结果表明过滤元件能去除气流中的大部分粉尘,实现对HTR-10一回路氦气中放射性石墨粉尘等悬浮物的高效过滤取样。     

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.     

The evolution of He+ irradiation-induced point defects and helium retention in nuclear graphite

The atomic-level study of point defect evolution in nuclear graphite is essential for a deep understanding of irradiation-induced property changes. The evolution of helium ion irradiation-induced point defects and helium retention in nuclear graphite ETU-10 and ETU-15 were studied by positron annihilation Doppler broadening (PADB) experiments and thermal desorption spectroscopy (TDS) measurements. The graphite samples were implanted with 10¹⁵, 10^16_, and 10¹⁷ cm^?2 of 200 keV He⁺ at operation temperatures below 373 K. Frenkel pairs were created during ion irradiation and they annihilated during annealing. Three stages of interstitial-monovacancy annihilation are suggested. At low temperatures, the initial annihilation would be refined only to the recombination of intimate metastable Frenkel pairs. When temperature increases, the annihilation would expand to a larger extent that isolate interstitials and vacancies annihilate with each other. In the case of high doses irradiation, vacancy clusters form at elevated temperatures. The retention and release of helium is tightly related to the evolution of the defects, especially the vacancies. The small over-pressured He-V clusters (He_nV) are thought to be the most possible form of helium retention under irradiation.     

氟盐冷却球床堆与HTR-10和MSRE的定量相似性分析

氟盐冷却球床堆是当前国际上一种新的研究堆型,尚无已经建造完成的反应堆,因此,选择相似且具有运行经验的反应堆作为基准题有助于堆芯核设计软件适用性分析。利用国际上常采用的相似性分析软件,可对熔盐实验堆(Molten Salt Reactor Experiment, MSRE)及10 MW高温气冷堆(10 MW high-temperature gas-cooled test reactor, HTR-10)与氟盐冷却球床堆的相似性进行分析,定量判断它们作为基准题的合理性。分析结果表明,MSRE和氟盐冷却球床堆的能谱峰位能量接近且堆内元素种类相近,二者相似程度较高;常温临界HTR-10和氟盐冷却球床堆冷却剂不同,且能谱峰位能量差异较大,二者相似程度较低。因此,MSRE是氟盐冷却球床堆中子物理设计软件较理想的基准题。     

Application of grey model on analyzing the passive natural circulation residual heat removal system of HTR-10

Using the grey correlation analysis, it can be concluded that the reactor pressure vessel wall temperature has the strongest effect on the passive residual heat removal system in HTR （High Temperature gas-cooled Reactor）, the chimney height takes the second place, and the influence of inlet air temperature of the chimney is the least. This conclusion is the same as that analyzed by the traditional method. According to the grey model theory, the GM（1,1） and GM（1, 3） model are built based on the inlet air temperature of chinmey, pressure vessel temperature and the chimney height. Then the effect of three factors on the heat removal power is studied in this paper. The model plays an import,ant role on data prediction, and is a new method for studying the heat removal power. The method can provide a new theoretical analysis to the passive residual heat removal system of HTR.     

HTR-10备用柴油发电机组的抗震鉴定

备用柴油发电机组是10MW高温气冷实验堆的重要I类抗震设备,采用在振动台上模拟地震试验方法进行抗震鉴定。本文详细介绍了抗震试验的方法和要求,并对试验结果进行了研究与分析。试验表明,机组能够满足HTR－10工程抗震的要求。     

HTR-10氦气流中石墨颗粒尺寸的估计

清华大学10MW高温气冷堆（HTR-10）采用石墨结构材料和石墨燃料元件,以及氦气冷却剂。由于结构部件的摩擦和磨损,反应堆一回路氦气流动中不可避免的带有石墨粉尘,这是反应堆设计中必须加以考虑的重要问题之一。本文根据凝并理论和颗粒学中的离散-分区模型（Discrete-Sectional Model,DSM）,建立了一种颗粒成长的计算方法,并对其进行了验证;同时运用该方法研究了HTR-10氦气流中石墨颗粒的发展情况,给出了氦气流中石墨颗粒在反应堆正常运行时的尺寸分布,并计算出石墨颗粒直径主要分布于10～20um,平均直径为12.9um。     

10MW高温气冷堆温度系数的测量和评价

在没有辅助热源的情况下．采用主氦风机循环和不大于500kW的核功率的联合加热方式加热堆芯并测量IOMW高温气冷堆的温度系数，用周期法测得石墨反射层平均温度从45．48℃到236．93℃过程中引起的反应性变化。同时，通过物理计算对测量结果进行评估．如果扣除氙毒和温度分布的影响．测量值和计算值是符合的。     

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.     

HTR-10技术规格书在线监督系统

10MW高温气冷堆(HTR-10)技术规格书在线监督系统以HTR-10仪表与控制系统提供的数据为基础,采用智能模拟运行人员行为的方法．实时分析判断反应堆系统和设备的工作状态是否满足技术规格书的要求,同时自动按照规定的频度完成部分定期试验和检查工作．对于不能自动完成的检查项目及时提醒运行人员,弥补了人工执行技术规格书时因工作量大、容易出现漏项的不足。本文介绍了HTR-10技术规格书在线监督系统的模型、组成模块、技术开发特点以及具体应用中的注意事项和方法。     

Scale removal oxidation behavior of metal in supercritical water modeled by cellular automaton

Corrosion and oxidation of structure material in supercritical water are specific and an important issue in the nuclear industry. A scale removal cellular automaton model was proposed to investigate the development of a continuous oxide layer of Inconel 625 in supercritical water at 24.8 MPa and 600 °C. This study presented influence of the reaction behavior of oxidation, scale removal effect, and transport ratio of oxygen and metal ions on the corrosion and oxidation process with different conditions. The formation of the spinel is simulated at mesoscopic level. The developed model is also mapped with the laboratory experimental data from a supercritical water loop.     

Thermodynamic Analysis of Thermal Efficiency of HTR-10 Hydrogen Production System

This paper presents thermodynamic analysis of the thermal efficiency of the 10 MW high temperature gas cooled reactor (HTR-10) hydrogen production system. The global reaction for the equilibrium reaction model is introduced. An analytical expression for the thermal efficiency is developed using the global reaction. For the specified temperature and pressure the thermal efficiency can be computed with the solution of the equilibrium. The investigation provides a more realistic limit for the efficiency of the nuclear hydrogen production system. The influence of the temperature, latent heat, steam-to-carbon ratio and pressure on the thermal efficiency is analyzed. Varying the temperature there is a maximum thermal efficiency for the specified pressure and steam-to-carbon ratio. The latent heat influences the thermal efficiency significantly, especially at the high temperature condition. Also varying the steam-to-carbon ratio there is a maximum thermal efficiency for the specified pressure and temperature. The process should be operated with high steam-to-carbon ratio to obtain maximum thermal efficiency when the reforming temperature is low and pressure is high. The maximum value is 68.9% within the range of the pressure greater than 1 MPa and steam-to-carbon ratio greater than 2. Comparison of theoretical results to experimental data is carried out.