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。     

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。     

用密度泛函理论研究氚在高温气冷堆核级石墨的吸附与解吸附行为

为研究氚在高温气冷堆核级石墨上的吸附和解吸附行为,本文利用密度泛函理论,采用氢原子代替氚原子的办法,通过理论计算得到了氚在高温气冷堆核级石墨上的结合能,通过模型分析得到了氚在高温气冷堆核级石墨上的吸附、解吸附机理与相应的份额,并得到HTR-10在20年寿期末各部分氚的累积量及事故工况下氚释放量的估计值。本文结果为研究估算高温气冷堆氚释放的机理提供了一条新思路。     

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。     

高温气冷堆用石墨材料的腐蚀实验与软件模拟

在高温气冷堆进水进空气事故下,空气和水蒸气会与堆内的石墨材料发生化学腐蚀反应,从而可能影响反应堆的安全。为研究高温气冷堆内石墨材料的氧化腐蚀特性,本文利用气相色谱法实验测量了IG-110石墨在不同温度和不同气体组分配比情况下的腐蚀速率及腐蚀产物,并利用THERMIX/REACT软件对整个石墨腐蚀过程进行了模拟。研究结果表明：反应温度对石墨腐蚀的影响最为显著,腐蚀速率随着温度的升高而增大,同时随着温度升高,CO与CO₂的含量比也逐渐增大。通过与实验结果对比分析,验证了THERMIX/REACT软件用于高温气冷堆安全分析的可靠性。     

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。     

HTR-10堆芯氧化模拟

对10 MW高温气冷堆(HTR-10)的堆芯模型进行简化,研究燃料元件在正常运行工况下的氧化情况,包括水蒸汽氧化及水蒸汽和氧气的共同氧化情况。结果表明,在燃料元件平均驻留期内石墨材料的水蒸汽腐蚀比较均匀,且主要发生在温度较高的底部;而氧气和水蒸汽对石墨材料的氧化则比较剧烈,底层燃料元件的石墨材料表面被腐蚀掉。     

高温气冷堆结构中的力学问题

本文着重介绍了高温气冷堆结构中的力学问题。由于高温气冷堆采用石墨作为主要结构材料,因而在高温、高辐照以及氧化气氛中石墨力学特性的研究是本文介绍的重点:诸如石墨在上述工作环境下的物理性质及力学性质,石墨的疲劳特性、石墨的应力分类及破坏准则等。对高温气冷堆压力容器(PCPV)应力分析中混凝土的徐变及开裂问题的研究,本文也做了介绍。     

核级石墨IG-110氧化模型研究

核级石墨IG-110在球床模块式高温气冷堆(HTR-PM)中作为结构材料被广泛应用。在正常运行工况及进气进水等严重事故工况下,核级石墨IG-110不可避免地与氧气等发生氧化反应,其氧化性能对反应堆的安全运行有着重要的作用。本文综合考虑核级石墨在氧化过程中局部孔隙率及氧化速率随失重率的变化,建立IG-110与氧气反应的氧化模型,得到氧化反应过程中各参数随空间和时间的变化。重点分析孔隙率变化和Knudsen扩散对有效扩散系数的影响,确定了在事故分析中较为适用的有效扩散系数形式。     

MELCOR程序在HTGR事故分析中的最新进展

MELCOR程序是美国NRC在安全评审中使用的一体化系统分析程序,早期主要用于轻水堆严重事故分析。近年来,该程序逐渐用于高温气冷堆的石墨腐蚀、裂变产物行为和石墨粉尘等物理现象方面的研究。本文介绍了在最新版本的MELCOR2．1程序中,针对高温气冷堆特点所进行的扩展和开发,以及MELCOR程序在高温气冷堆（HTGR）事故分析中的计算流程。     

The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.     

Oxidation experiments and theoretical examinations on graphite materials relevant for the PBMR

Graphite oxidation due to gas impurities in normal operation and ingressing oxidants in accidents plays a key role in the material and safety behaviour of HTRs. An overview is presented of the theoretical background concerning graphite oxidation, mainly in regimes I and II. Some differences between the classical oxidation model, based on effective diffusivity and chemical reaction on the inner graphite surface, are discussed. These differences may be due to the complex pore system in graphite, which cannot be approximated by one single diffusivity. Based on these theoretical results, a procedure for measurements on candidate graphites to be used in PBMR is proposed. Regime I measurements are selected for material characterization because of the strong sensitivity to chemical influences. First results measured in air at 650-750 °C at the Graphite Oxidation Laboratory, GOLab, Research Centre Jülich, are outlined. Graphites examined so far are the SGL grades NBG-10 and NBG-18. Whereas NBG-10 is significantly more oxidation resistant for all specimens and at all temperatures than the former German nuclear graphite V483T5, taken as a standard, the scatter of oxidation rates of NBG-18 is even larger, but is on average also satisfactory. In contrast to the classical model, preliminary low-temperature oxidation experiments on NBG-10 reveal a significant rate dependence on specimen size. Additional experiments in regime I and in regime II are proposed for PBMR graphites, as those for clarification of the deviations to the classical oxidation model. The latter probably requires a broader discussion in the graphite community.     

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。     

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。     

Large-scale numerical simulation of mechanical and thermal properties of nuclear graphite using a microstructure-based model

The mechanical and thermal properties of nuclear graphite depend strongly on the microstructures. In this paper, a large-scale three-dimensional boundary element model is presented to study the relationships between the bulk effective properties and microstructure changes in nuclear graphite. Acceleration of the associated boundary element method (BEM) is achieved by use of a fast multipole method (FMM) in allowing large-scale numerical simulations of the model containing up to several hundred micro-structural pores to be performed on one desktop computer. The effects of several key micro-structural parameters such as the pore aspect ratio and the fractional porosity on the bulk mechanical and thermal properties of nuclear graphite are evaluated. The numerical results are compared with some experimental data due to oxidation and good agreement is observed. It is demonstrated that the presented method is potential for fundamental understanding of the bulk properties of nuclear graphite from micro-structural views.     

Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basic phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. We plan a scaled-model test and a full-scale model test for the LOVE.     

Behavior of graphite in nuclear reactor stacks

The results of a study of radiation effects in graphite and the influence of those effects on the design of graphite stacks for nuclear reactors are discussed in this article. Several of the manifestations of these effects may lead to serious complications in reactor performance. Measures used to avert such complications are considered. A unified approach to the physical nature of the radiation effects in graphite is suggested for a broad range of elevated temperatures. The problem of preventing oxidation of graphite is approached in the light of the high temperatures prevailing in the graphite stacks of reactors of recent design.