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。     

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。     

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.     

V型凹口对液态熔盐堆用核石墨构件应力分析的影响

本文利用了一个根据球床模块堆(Pebble Bed Modular Reactor,PBMR)用核石墨材料辐照性能数据编写的用户自定义材料模型(User defined Material model,UMAT),按照美国橡树岭国家实验室(Oak Ridge National Laboratory,ORNL)的液态燃料熔盐试验堆(Molten Salt Reactor Experiment,MSRE)用核石墨构件尺寸,为钍基熔盐堆(Thorium-based Molten Salt Reactor,TMSR)设计了一款方型核石墨构件。利用新编UMAT对该核石墨构件进行了初步的应力分析。分析结果表明,在没有预制裂纹的情况下辐照梯度越大核石墨构件中心区域最大主应力值越大,构件的断裂位置可能出现在构件中心位置处;对于有V型凹口预制裂纹的情况,应力集中部位均出现在预制裂纹尖端附近,这将可能导致裂纹尖端附近出现裂纹扩展,从而引起构件断裂失效。     

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。     

载荷对HTR-PM高温气冷堆用石墨球摩损性能的影响

采用摩擦磨损试验机研究了载荷对HTR-PM高温气冷堆用石墨球在氦气气氛下的摩损性能影响,并利用光学显微镜分析了磨损表面形貌,利用激光粒度仪和扫描电镜分析了石墨粉尘的粒度和形貌。结果表明:在氦气气氛下,摩擦系数随着载荷的增加呈现减小的趋势;产生的石墨粉尘主要呈片状或絮状体,有较强的粘附性,粒度随着载荷的增加而增加;不同载荷下,随着时间的增加,石墨球的磨损率不断减小,最终趋于一个稳定值。     

犬骨型核石墨四点弯曲实验与模拟

核石墨是熔盐堆的关键材料之一,断裂性能是核石墨的重要属性之一。首先通过四点弯曲实验测量了犬骨型核石墨的断裂载荷,观察裂纹扩展路径再运用扩展有限单元法(Extended finite element method,XFEM)对这一实验过程进行了模拟。模拟得到的裂纹扩展路径和断裂实验结果有很好的一致性,证明利用XFEM可以准确地模拟核石墨的断裂过程。同时确定了适用于核石墨的断裂准则。     

HTR-PM高温气冷示范堆堆芯石墨粉尘产生量估算

在球床式高温气冷堆堆芯内,影响石墨球摩擦磨损率的关键条件为载荷与温度。此前,中国辐射防护研究院研究了载荷对石墨球摩擦磨损性能的影响,得到了石墨球磨损率与载荷的关系。本文在此基础上进一步研究了温度对石墨球磨损率的影响,通过拟合得到了石墨球磨损率与石墨球所受载荷、温度之间的关系式,结合HTR-PM高温气冷示范堆内燃料元件所受载荷和温度的分布情况,计算得出石墨球之间摩擦产生的石墨粉尘量约为14.01 g/d(5.1 kg/a)。     

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。     

HTR-10堆芯球流运动的唯象学DEM模拟

清华大学研发的10 MW高温气冷堆（HTR-10）是国际上重要的先进实验反应堆,球流运动的研究具有基础性地位。通过唯象的方法对HTR-10堆芯的球流运动进行了离散元数值模拟,通过已由实验验证的计算程序,采用与HTR-10堆芯1∶1的计算模型,计算了27 000个元件单元的运动,包括不同摩擦系数f和不同底部锥角A下的球流运动。结果表明：在HTR-10堆芯设计条件下,球流运动较均匀,堆芯底部不存在滞留区;f越大或A越大,堆芯球流越均匀,表现出更好的整体性向下运动;当f达到0.8上限时,HTR-10堆芯球流依然保持了整体性运动,底部无任何被滞留的球。本工作对进一步优化球床式高温气冷堆堆芯设计具有重要意义。