In situ transmission electron microscopy of electron-beam induced damage process in nuclear grade graphite

Atomic level processes involved in the swelling and crack-closing in nuclear grade graphite under electron irradiation have been observed in real-time using transmission electron microscopy. Noise-filtered lattice images show the formation of vacancy loops, interstitial loops and resulting dislocations with unprecedented clarity. The dislocation dipoles formed via vacancy loops were found to undergo climb resulting in extra basal planes. Concurrent EELS studies showed a reduction in the atomic density because of the breakage of hexagonal carbon rings. The formation of new basal planes via dislocation climb in addition to the bending/breaking of basal planes leads to swelling and closing of micro-cracks.     

Buckle, ruck and tuck: A proposed new model for the response of graphite to neutron irradiation

The default theory of radiation damage in graphite invokes Frenkel pair formation as the principal cause of physical property changes. We set out its inadequacies and present two new mechanisms that contribute to a better account for changes in dimension and stored energy. Damage depends on the substrate temperature, undergoing a change at approximately 250 °C. Below this temperature particle radiation imparts a permanent, nano-buckling to the layers. Above it, layers fold, forming what we describe as a ruck and tuck defect. We present first principles and molecular mechanics calculations of energies and structures to support these claims. Necessarily we extend the dislocation theory of layered materials. We cite good experimental evidence for these features from the literature on radiation damage in graphite.     

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。     

The nucleation and early growth of interstitial dislocation loops in irradiated materials

A hierarchy of rate equations is solved to describe the homogeneous nucleation of interstitial dislocation loops in irradiated materials. Calculations for graphite and M316 stainless steel have been performed. The concept of nucleation time is examined and a procedure is adopted which gives a useful criterion for defining the end of the nucleation period. Calculations have been performed which demonstrate the effects of temperature, dose rate and network-dislocation density on the nucleation and final concentration of interstitial loops. The assumption that di-interstitial atom pairs are stable against thermal dissociation is examined and found to be appropriate for the conditions used in this work.     

加速器质谱14C样品制备中Zn-TiH2法石墨产率探讨

用于加速器质谱仪(accelerator mass spectrometry, AMS)测定¹⁴C年龄的石墨靶制备方法有很多种,提高石墨产率对优化石墨靶制备流程和改善石墨靶性能具有重要意义。本文以Zn-TiH₂法制备石墨靶为例,结合相关研究,证明石墨产率与同位素分馏之间的关系;对比石墨产率的不同测算方法并提出改进建议;对实验过程中影响石墨产率的因素进行探讨,对Zn-TiH₂法制备石墨靶的最优实验条件进行简要总结。     

Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

Continuum modeling of localized deformation in irradiated bcc materials

A continuum constitutive crystal plasticity framework is implemented to model the post-irradiation tensile behavior of bcc structural materials, accounting for localized deformation due to the formation of dislocation channels. Both the mechanical response and deformed microstructure of the material are modeled for quasi-static tensile loading. The latter is studied to identify the stages of dislocation channel formation during localization, specifically with respect to the evolution of dislocations and irradiation-induced defects. Parametric studies of the cross-slip and flow softening (due to annihilation of irradiation-induced defects) models are performed to study their effects on the localization behavior. Results are compared to available experimental data.     

核级石墨氧化腐蚀模型研究

高温气冷堆内应用到大量核级石墨材料,对其长期氧化腐蚀行为进行研究至关重要。文章建立了综合考虑石墨内部孔隙率变化及失重率影响的石墨氧化模型,对气体在石墨内部的瞬时氧化腐蚀情况进行了模拟计算。提出氧化深度的概念,研究发现反应温度越高,反应气体在石墨内部的氧化深度越小;并与实验结果及其他模型的计算结果进行了对比,验证了模型的有效性。     

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。     

真空石墨加热器温度场数值模拟与分析

石墨加热器是测量球床堆芯等效导热系数实验的关键部件,加热器温度场对系统安全及数据准确性有重要影响。本文基于Fluent计算平台,分别采用DTRM模型、P1模型、ROSSELAND模型、DO模型对真空保护环境下的石墨加热器温度场进行数值模拟,确定适合真空保护石墨加热器温度场的计算方法并讨论石墨导热系数、表面发射率对温度场分布的影响。比较分析表明：DO模型计算得到的温度分布较为接近真实情况,导热系数小于35W/(m•K)时,最高温度对其敏感;导热系数大于35W/(m•K)时,其对加热体最高温度影响较小,最高温度较为稳定。     

SCANNING TUNNELING MICROSCOPY OF DNA

Uncoated DNA deposited on graphite surface was imaged in air with the scanning tunneling microscope (STM). The images revealed a helical pitch of 3.6 nm whereas the average width was 2.3nm. Periodic alternation of major and minor grooves were resolved with high resolution. Our STM images were in reasonable agreement with the B-form DNA structural model derived from X-ray diffraction studies. It was concluded that the STM could be useful for local structural studies of biomolecules.     

孔隙率对核级石墨氧化的影响

核级石墨在高温气冷堆中作为结构材料、慢化材料和反射层材料等被广泛应用,其氧化性能对高温气冷堆在进水或进气事故下材料的腐蚀行为有重要影响。初始孔隙率分布及孔隙率在氧化过程中的变化均对石墨氧化造成影响。本文以核级石墨IG-110、H-451、NBG-18和A3-3为例,以直径为6 cm的石墨球为研究对象,在一维瞬态氧化模型的基础上,分析了初始孔隙率分别服从均匀分布、正态分布和对数正态分布时对石墨氧化的影响。从模型简化和高温气冷堆安全分析角度保守考虑,建立石墨氧化模型时,核级石墨初始孔隙率可取均匀分布,此时石墨的整体失重率最大。     

Prediction of fuel channel—graphite gas-gap behaviour in RBMK reactors

To maintain thermal contact between the fuel assembly and the graphite moderator, RBMK design reactors employ graphite split rings, which are alternatively tight on the pressure tube or tight on the graphite brick central bore. The split in the graphite rings allows a helium/nitrogen gas mixture to flow up the fuel channel. This prevents oxidation of the graphite and can be sampled to detect pressure tube leaks. The initial clearance between the rings and pressure tube or graphite brick is approximately 2.7 mm (1.35 mm each side). Due to material property changes of the pressure tubes and graphite during operation of the reactor, the size of the clearance between the rings and the pressure tube/brick, called the “gas-gap”, varies. Closure of these gaps has been identified as a possible safety case issue by reactor designers and by independent reviews carried out as part of TACIS reviews and as part of the Ignalina Safety Analysis Report. The reasons for this are that gas-gap closure would cause the pressure tube to be tightly gripped by the graphite bricks via the split rings, which could lead to:

• Extra loading on the upper pressure tube zirconium/steel transition joint, particularly during shut down and emergency transients.

• Splitting of the graphite brick, leading to loss of thermal contact between the pressure tube and graphite. As approximately 5.6% of the heat in graphite-moderated reactor is generated within the moderator through neutron and gamma-heating, loss of thermal contact would result in higher graphite temperatures, accelerating the rate of graphite expansion and hence increasing the loading of the core radial restraint.

• Graphite debris may become lodged in inter-brick gaps, leading to increased axial pressure tube loading during shut down and emergency transients.

The authors have carried out deterministic assessments based on the Ignalina RBMK-1500 reactors in Lithuania, modelling the behaviour of the graphite under irradiation and have predicted graphite bore diameter changes that are in good agreement with the measurements of graphite bore diameters taken at Ignalina Nuclear Power Plant (NPP). A probabilistic model has been developed using the actual results of the deterministic calculations with non-linear graphite behaviour. Statistical analysis of the measurements of tube and graphite diameters taken from Units 1 and 2 at Ignalina NPP has been carried out. Further work has been carried out to try to determine the uncertainty inherent in the predictions of the gas-gap closure from the calculations. The overall objective of the studies is to aid prediction of the gas-gap closure process, and help to identify a suitable monitoring strategy for gas-gap closure that could be used for any RBMK reactor.     

高温气冷堆核级石墨相关问题研究

本文通过对石墨在高温气冷堆中的运行环境进行了分析,研究了在石墨堆内构件设计中的关键问题和在高温气冷堆单个模块及其未来发展中核级石墨的需求。从原料、成型及中子辐照等角度分析了核级石墨国产化研究方向。根据核级石墨目前的研发形势,进行了风险问题分析。     

针对细颗粒石墨的改进概率评价方法研究

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4代高温气冷堆中。由于石墨材料强度分散,与常用的确定论评价方法相比,概率论方法评价其失效更为合适。本文通过有限元软件ABAQUS用户子程序开发了石墨构件失效概率分析模型,采用该模型研究了Hindley模型对细颗粒及国产石墨的适用性,在此基础上提出改进的失效概率计算模型,并通过试验数据加以验证。结果表明,Hindley模型过于保守,改进模型则很好地吻合了试验数据,其结果更为合理,为国产石墨在核反应堆中的应用提供参考依据。     

Acoustic emission caused by reversed plastic deformation in polycrystalline graphite

The reversed plastic deformation in polycrystalline graphite grade IG-11 during unloading has been verified using the acoustic emission technique. The change in root mean square (RMS) voltage for continuous acoustic emission was detected as IG-11 graphite was plastically deformed. In Pyroid graphite, the RMS voltage for continuous acoustic emission was unchanged around zero level as long as the graphite was elastically deformed. The shape of plastic strain rate-total strain curve in IG-11 graphite was relatively in good agreement with that of RMS voltage-total strain curve under loading. The cause of continuous acoustic emission was attributed to the plastic deformation for IG-11 graphite. The reversed plastic deformation was shown by using acoustic emission technique during unloading in compressive and tensile tests as we expected. The mechanism of the reversed plastic deformation can be explained using the bicrystal model of polycrystalline graphite. Upon reimposing load, the Kaiser effect is not observed in polycrystalline graphite, which resulted from the reversed plastic deformation.     

石墨垫片密封泄漏率计算方法研究

为完善核级主设备密封分析及设计方法,基于稳压器人孔密封结构建立了密封数值分析模型,对石墨垫片密封接触应力进行了分析研究;结合平行圆板流动模型和多孔介质渗流模型建立了石墨垫片密封质量泄漏率理论预测模型;基于理论预测模型计算了设计工况、试验工况和启动瞬态工况下的质量泄漏率,对主要影响参数进行了分析和讨论。研究结果表明,石墨垫片密封接触应力沿周向分布较为均匀,而石墨环沿径向的中间区域接触应力值略低于石墨环两侧;在温度和压力上升瞬态中,密封接触应力随时间呈现出下降的规律,密封质量泄漏率与接触应力呈负相关,增大密封接触应力可以降低质量泄漏率,但降低效率逐渐减小,减小粗糙度可以显著降低质量泄漏率。本文分析方法可为核级主设备密封泄漏率分析和紧密度评价提供重要参考。     

高温气冷堆用石墨材料氧化行为研究进展

核级石墨是高温气冷堆重要的慢化剂、反射层和结构材料,其氧化腐蚀性能对反应堆安全运行至关重要,因此已成为核材料学科的研究热点之一。本文综述了国内外在核级石墨氧化腐蚀领域的研究现状,总结了核石墨氧化的化学动力学模型、失重率影响因子模型以及模拟计算模型,提出了高温气冷堆用石墨材料氧化腐蚀的研究方向。     

CONFORMATION STUDY OF THE LARGE FRAGMENT OF E. COLI DNA POLYMERASE I BY STM

The large fragment of E.coli DNA polymerase I is imaged by scanning tunneling microscope. The specimen is deposited on highly oriented pyrolytic graphite surface, and then covered with pure paraffin oil in order to maintain hydration of the molecules. Images of the enzyme reveal an ellipsoid shape of 5.5-6.0nm wide and 7.0 -7.5 nm long. The conformation of the enzyme is in agreement with the model derived from X- ray crystallography studies.     

Modelling of the effect of dislocation channel on intergranular microcrack nucleation in pre-irradiated austenitic stainless steels during low strain rate tensile loading

In the present article, the effect of dislocation channel on intergranular microcrack nucleation during the tensile deformation of pre-irradiated austenitic stainless steels is studied. Because several slip planes are activated within the dislocation channel, the simple dislocation pile-up model seems not well suited to predict grain boundary stress field. Finite element computations, using crystal plasticity laws and meshes including a channel of finite thickness, are also performed in order to study the effect of some microstructural characteristics on grain boundary stress field. Numerical results show that: the thickness and the length of the dislocation channel influence strongly the grain boundary normal stress field. The grain boundary orientation with respect the stress axis does not affect so much the grain boundary normal stresses close to the dislocation channel. On the contrary far away the dislocation channel, the grain boundary stress field depends on the grain boundary orientation. Based on these numerical results, an analytical model is proposed to predict grain boundary stress fields. It is valuable for large ranges of dislocation channel thickness, length as well as applied stress. Then, a macroscopic microcrack nucleation criterion is deduced based on the elastic-brittle Griffith model. The proposed criterion predicts correctly the influence of grain boundary characteristics (low-angle boundaries (LABs), non-coincident site lattice (non-CSL) high-angle boundaries (HABs), special grain boundaries (GBs)) on intergranular microcrack nucleation and the macroscopic tensile stress required for grain boundary microcrack nucleation for pre-irradiated austenitic stainless steels deformed in argon environment. The criterion based on a dislocation pile-up model (Smith and Barnby) underestimates strongly the nucleation stress. These results confirm that pile-up models are not well suited to predict microcrack nucleation stress in the case of dislocation channels impacting grain boundary. The proposed criterion is applied to the prediction of the IASCC macroscopic nucleation stress for pre-irradiated material tested in PWR environment and the predictions are discussed with respect to experimental data. Finally, the limitations of the continuum modelling are discussed.