不确定性分析技术在某环保配套工程安全评价中的应用研究

为解决放射性废物处置安全评价结果受认知水平和长时间尺度等因素引起的不确定性问题，基于某环保配套工程项目开展确定性模型安全评价，然后基于概率论方法、拉丁超立方抽样技术生成1 000个样本，开展不确定性分析和灵敏度分析工作。结果表明确定性模型相对于不确定性模型，对近场释放率计算结果偏高，对地质圈释放率计算结果良好，对Mo-93、Ni-59产生的照射剂量估算较好，对I-129产生的照射剂量代表性中等偏下。灵敏度分析结果表明工程屏障中混凝土相关参数(如：混凝土有效扩散系数、混凝土中核素分配系数、混凝土屏障厚度)、废物浸出率、初始活度和核素在地质圈的分配系数为模型释放率峰值主要影响参数，可为后续工程设计与野外调查遴选出重点参数。     

高放废物地质处置库热学分析的参数敏感性研究

应用FLAC3D软件建立高放废物地质处置库热学分析的简化计算模型,选择影响处置库温度场的包括材料热学参数、几何参数以及时间参数在内的16个关键参数,以膨润土内表面峰值温度(该物理量是高放废物地质处置库热学设计计算中作为温度准则的物理量)为参数敏感性分析的目标物理量,通过热学计算开展参数敏感性分析。在参数敏感性分析中,将参数敏感程度划分为高、中、低三等。分析表明:4个参数(膨润土导热系数、膨润土厚度、围岩导热系数、高放废物中间贮存时间)为高敏感度参数,2个参数(散热材料厚度、回填材料厚度)为中度敏感性参数,其它10个参数(高放玻璃固化废物体、外包装容器、散热材料、回填材料的导热系数与比热,以及膨润土与围岩的比热)为低敏感度参数。通过分析可以得到如下结论:在设计高放废物地质处置库时,对膨润土及围岩导热系数的测试应力求准确,对测试结果数据认真分析,确保为设计计算提供合理的输入参数;在确保膨润土满足工艺要求功能的前提下,宜尽量减小膨润土的厚度;按照本文热学分析模型初步估算,我国高放废物至少需要中间贮存20 a以上。     

国内外核废物处置库近场温度场模拟预测

核废物处置后因所含的放射性核素衰变而产生的衰变热通过传导、对流以及辐射等方式从废物体向外传递,从而引起废物罐体、缓冲材料及近场围岩温度升高,导致废物体至近场围岩之间形成温度梯度。温度梯度随着时间的延续而变化,最终会影响地下水系统和核素迁移。本文对一些国家的处置库温度预测模式进行了调研,对源项、处置库模型简化、热传递数学模型和模拟结果做了初步总结,为我国拟建处置库的温度场预测提出了建议。     

膨润土-砂混合物缓冲层轴向热传导性研究

膨润土-砂混合物作为高放废物处置库缓冲材料,在高放废物衰变释热作用下,其物理力学性能对处置库的稳定和安全性具有重要影响。本研究采用自行设计的装置,对按比例缩小后的不同干密度、含水率、掺砂率试样进行热传导模拟试验,并对缓冲层热-力耦合过程进行数值模拟分析,得到了缓冲层温度、应力和应变的变化及分布情况,重点分析了温度的影响。结果表明,增大试样干密度、含水率和掺砂率均可提高其导热性,应变也随之增大,应力受温度影响较早达到平衡;缓冲层靠近热源的位置温度、应力和应变最大,沿轴向方向递减,初始时刻变化明显。     

大气条件下高庙子钠基膨润土化学缓冲特性研究

深地质处置目前被国际上公认为是处置高放废物的最有效可行的方法。我国采用多重工程屏障系统和适宜的地质体共同作用来确保与生物圈的安全隔离。缓冲材料是高放废物重要的工程屏障材料之一,我国选用高庙子钠基膨润土作为缓冲材料的基础材料。膨润土作为缓冲材料的一个重要性能表现为缓冲孔隙水的化学变化。介绍了GMZ-1钠基膨润土大气条件下与蒸馏水的反应试验,并对试验结果进行了讨论。批式试验反应溶液中钠离子来源于钠基膨润土层间阳离子和矿物溶解,镁离子来源于钠基膨润土层间阳离子,钾离子和钙离子来源于矿物溶解,相关研究认识对于高放废物处置库近场核素迁移研究和评价工程屏障的长期稳定性具有重要意义。     

中低放废物近地表处置库源项释放模式及模拟计算结果

为了计算中低放废物近地表处置库源项释放速率，本文以某核电站处置场的概念设计为例，建立了一个简单而比较完整的源项释放模式。它包括水入渗模式、处置库混凝土顶盖的破损模式，金属桶腐蚀模式、核素从水泥固化体中的浸出释放模式及浸出核素在回填材料的中的迁移模式。     

高放废物地质处置近场核素迁移计算

以我国高放废物地质处置初步概念设计为背景,分析了高放废物地质处置系统中近场核素迁移的基本途径。初步建立了放射性核素近场迁移的概念模型和数学模型,并采用库室模型方法,利用Goldsim通用软件模拟了在参考景象下放射性核素从近场的释放迁移,给出核素从近场的释放率结果。采用日本的概念设计和参数值,利用建立的模型进行了计算,与日本原子能机构JAEA给出的H12报告(使用MESHNOTE程序计算)的结果进行了比较,两者基本吻合。最后提出了需要进一步研究的内容。     

Ca^(2+)盐溶液在高庙子膨润土中的扩散研究北大核心CSCD

高放废物地质处置库近场地下水可能会对处置库内的屏障体系产生影响,降低处置库的安全稳定。为研究地下水中盐离子在处置库内缓冲回填体系的扩散规律,本文开展了静态无外荷载条件下内蒙古高庙子(GMZ)膨润土在Ca^(2+)盐溶液中自发渗吸的吸附扩散室内试验。从土的微观结构和经典扩散理论对Ca^(2+)在不同干密度和初始饱和度的膨润土试样中的自发扩散规律进行了分析。研究结果表明,在膨润土初始饱和度相同的情况下,试样阻滞系数随其干密度增加而增大,此时Ca^(2+)的扩散能力减弱;当膨润土干密度相同时,随着初始饱和度的增加基质吸力作用减弱,阻滞系数减小,Ca^(2+)的扩散能力减弱。     

甘肃北山高放废物处置库预选场址远场环境影响模拟评价

以甘肃北山高放废物处置库预选场址为评价对象,收集该预选场址的环境资料,采用Ecolego软件对该场址远场环境进行评价。远场环境评价时考虑的公众受照途径为黑河流域水灌溉和饮用造成的公众辐射剂量。经评价,个人所受剂量在高放废物处置库关闭后随时间逐渐增大,在关闭后1.5×10~6 a左右达到剂量率峰值4.66×10~(-11) Sv/a,此时,关键核素为~(135)Cs,剂量贡献值为99.1%。     

中国高放废物地质处置研究进展:1985～2004

如何安全处置高放废物是核工业可持续发展面临的挑战性问题。我国的高放废物深地质处置研究从1985年开始，提出的计划目标是：于21世纪中叶建成我国高放废物地质处置库，处置的对象是玻璃固化块、超铀废物和部分乏燃料，处置库为竖井一坑道型，候选围岩为花岗岩，位于饱和带中。在1985～2004的20a中，我国高放废物地质处置研究取得了进展，已确定我国高放废物最终处置走“深地质处置”，并且是“三步曲”式的技术路线，即处置库选址和场址评价一地下实验室一处置库。经过全国筛选对比，已初步选定甘肃北山地区为重点预选区，该区地处戈壁，地壳结构完整，地壳稳定，人烟稀少，地质条件和水文地质条件均有利。20世纪90年代初期，开展了地下实验室的选址工作，初步选择了北京郊区2处地点为我国高放废物地质处置“普通地下实验室”的场址。已确定使用膨润土作为处置库的回填材料，并初步确定内蒙古高庙子膨润土为我国高放废物处置库的首选缓冲回填材料。对膨润土的矿物学、岩土力学、物理力学性质和热学性质进行了研究。已获得一批放射性核素(主要是Np、Pu、Tc)在北山花岗岩和膨润土上的吸附分配比、扩散系数和弥散系数等参数，建立了低氧手套箱和模拟处置库温度、压力和氧化一还原条件的小型实验装置。高放废物中的关键核素的化学行为研究也取得进展。花岗岩接触带核素迁移、铀矿床中超铀元素迁移、青铜器腐蚀等天然类比研究取得了成果。还开展了高放废物地质处置系统总性能评价源项和生物圈模式的调研。概念设计研究仅在20世纪90年代初开展了部分研究。从1999年开始，与国际原子能机构开展了2期高放废物地质处置技术合作项目，极大地提高了我们的技术水平。20a的科研工作为我国在21世纪完成高放废物地质处置奠定了一定基础。     

多因素耦合作用下缓冲材料对铀的长期阻滞效应数值模拟北大核心CSCD

近场环境条件下核素在缓冲材料中的迁移扩散受控于温度场、渗流场、膨胀应力场和化学吸附场的耦合作用,其对核素的阻滞特性将影响到核素随地下水向处置库围岩迁移并返回生物圈的能力,开展多因素耦合作用下缓冲材料对铀的长期阻滞效应研究,对地质处置库的长期安全性评价具有重要的意义。本研究基于混合物理论、连续介质理论、质量守恒、动量守恒、能量守恒及溶质扩散的Fick定律,推导出饱和缓冲材料中核素迁移扩散的热-水-力-化耦合控制方程,并借助于COMSOL Multiphysics软件的直接全耦合求解优势,以自主研制的缓冲材料长期阻滞性能Mock-up实验装置为几何模型,采用内置接口和添加热-水-力-化耦合控制方程中的耦合项作为源项相结合方式,实现了多物理场耦合作用下铀在饱和缓冲材料中迁移扩散行为的直接耦合分析,其长期阻滞特性数值模拟结果表明:初期阶段铀在缓冲材料中迁移扩散较缓慢,迁移距离随时间增幅在1 m左右;中后期阶段,随缓冲材料对铀的吸附容量逐渐趋于饱和后,其迁移距离较初期阶段增加更为明显,迁移距离随时间增幅为3 m左右。多因素耦合下核素在饱和缓冲材料中迁移扩散的热-水-力-化耦合控制方程构建、求解及长期阻滞性能模拟研究的方法,能够为我国高放废物深地质处置库地下实验室开展1∶1工程尺度的工程屏障设计与安全性能评价提供技术参考。     

Preliminary and Scoping Analysis for Uncertainty of HLW Geologic Repository Performance with Multiple Canister Repository Model by Earth Simulator

An uncertainty analysis of repository performance has been made by the VR code, which incorporates interference effects of multiple canisters in a repository. A problem of the previous study with the VR code is that the number of connected canisters was determined arbitrarily. In this study, first, the probability distribution functions for the number of canisters connected by the flow-bearing fracture clusters have been determined by the FFDF code, and then uncertainty associated with the peak fractional release rate of ²³⁷Np to the far field resulting from uncertainties with the buffer sorption coefficient, solubility, and the number of connected canisters has been numerically evaluated by the Latin-Hypercube Sampling method. The effects of uncertainty with the total number of connected canisters become less important as the number increases because the radionuclide concentration saturates in downstream compartments. Uncertainty with the buffer sorption and solubility shows an important contribution to that with the nuclide release rate.     

核素释放率计算方法的比较

核素释放率是尾矿库及核废物处置场地环境影响评价的基础上，本文比较了在环境影响评价中常见的两种计算核素率的方法，对它们的特点进行了评述。     

Review of waste glass corrosion and associated radionuclide release as a part of safety assesment of entire disposal system

Current knowledge on high-level nuclear waste glass corrosion is summarized, and remaining problems are discussed for meaningful predictions of the glass corrosion and associated radionuclide release as a part of safety assessment of entire disposal system. In recent years, much progress has been made in understanding the mechanism of waste glass corrosion in aqueous environments. Glass corrosion models based on the mechanism have been developed for predicting the long-term glass performance, and they are incorporated as part of radionuclide source term in safety assessments of the disposal system. However, these results have not yet allowed meaningful predictions for the long-term release of individual radionuclides from the glass in repository environments, because mechanism of the long-term glass corrosion has not been fully understood and solubilities of actinoids and fission products under disposal conditions are rather uncertain. In addition, the most serious problem is that the effects of various reactions and interactions occurring in the engineered barrier system, such as corrosion of overpack, alteration of backfill and chemical interactions of the released glass constituents with them have not been fully coupled with the glass performance. These reactions may be dominant processes controlling the glass corrosion and associated radionuclide release for the long-term. For the meaningful predictions, we must evaluate the waste glass performance in combination with the effects of various reactions and interactions occurring in the engineered barrier system on the basis of fully understanding of the chemical and geochemical mechanisms.     

Geochemically Based Safety Assessment

Performance assessment (PA) for nuclear waste disposal based on transport processes frequently neglects significant safety factors by overestimation of radionuclide mobilization and underestimation of radionuclide retention processes. To include well understood geochemical knowledge into PA, the quasi-closed system approach (QCS) was developed. The QCS approach is described and applied to a LLW disposal in a German salt mine with respect to the disposed waste forms, geo-engineered barriers, and backfill strategies. The geochemical tools and the thermodynamic database for modelling highly concentrated salt systems are discussed. Applications are demonstrated which cover the long-term geochemical environment in the disposal caverns, the optimization of buffer materials, radionuclide retention, and the overall robustness of the approach. Also the effect of a potential solution exchange between different emplacement caverns is investigated. It is shown that the QCS approach provides essential data concerning the long-term geochemistry and related radionuclide concentrations to be used in PA and safety analysis.     

Development of compartment models with Markov-chain processes for radionuclide transport in repository region

This paper presents a new radionuclide transport model for performance assessment and design of a geologic repository for high-level radioactive waste. The model uses compartmentalization of a model space and a Markov-chain process to describe the transport. The model space is divided into an array of compartments, among which a transition probability matrix describes radionuclide transport. While similar to the finite-difference method, it has several advantages such as flexibility to include various types of transport processes and reactions due to probabilistic interpretation, and higher-order accuracy resulting from direct formulation in a discrete-time frame.We demonstrated application of this model with a hypothetical repository in porous rock formation. First we calculated a three-dimensional steady-state heterogeneous groundwater flow field numerically by the finite-element method. The transition probability matrix was constructed based on the flow field and hydraulic dispersion coefficient. The present approach has been found to be effective in modeling radionuclide transport at a repository scale while taking into account the effects of change in hydraulic properties on the repository performance. Numerical exploration results indicate that engineered barrier configuration and material degradation have substantial effects on radionuclide release from the repository.     

Analyses of disposal efficiency based on nuclear spent fuel cooling time and disposal tunnel/pit spacing for the design of a geological repository

The purpose of deep geological disposal of high-level radioactive waste (HLW) including nuclear spent fuels is to isolate and to inhibit the release of radioactive material for a long time so that its toxicity does not affect the biosphere. The main requirement for the HLW repository design is to keep the buffer temperature below 100 °C in order to maintain the integrity of the engineered barrier system. The cooling time of the spent fuels discharged from nuclear power plants is the key consideration factor for the efficiency and economic feasibility of such a repository. We analyze the spacing of the disposal tunnels and pits, the disposal area and the uranium density for the deep geological repository layout to satisfy the thermal requirement of the disposal system. To do this, thermal stability analyses of a disposal system have been performed using varying spent fuel cooling times and spacing of the disposal tunnels and pits. The results show that the time to reach the maximum temperature within the design limit of the temperature in the disposal site is likely to be shortened as the cooling time of the spent fuel becomes shorter. Also it seems that controlling the disposal pit spacing is considered more advantageous than controlling the disposal tunnel spacing to meet the allowable thermal criteria in the repository from thermal and economical points of view. The results of these analyses can be used for a deep geological repository design and detailed analyses with exact site characteristics data will reduce the uncertainty of the results.     

Estimates of Parameter and Scenario Uncertainties in Shallow-Land Disposal of Uranium Wastes Using Deterministic and Probabilistic Safety Assessment Models

Safety and uncertainty analyses for the shallow-land disposal of uranium wastes were performed using the deterministic and probabilistic safety assessment models. The analyses for uranium accumulated with 4.5% enrichment show that the doses in residence scenario are of great importance in the safety assessment owing to the influence of daughters built up by uranium decay chain. The dose in residence scenario is sensitive to the release condition of radionuclides from the facilities over long-term period. The parameter uncertainties for the important pathways in residence scenario are estimated from the probabilistic analyses using the statistical methodology. The uncertainty analysis indicates that the influence of parameter uncertainty is the most remarkable in the estimation for the inhalation of radon gas with residence. The parameter importance in each exposure pathway is estimated from using the partial rank correlation coefficients (PRCCs) between variable parameters and the evaluated doses. The important parameters identified by the PRCCs are depth of intrusion, infiltration rate, thickness of covered soil, diffusion coefficient of radon in soil etc. for the inhalation exposure of radon.     

集成缓冲材料对锶的长期阻滞效应

缓冲材料作为高放废物处置库中多重屏障体系的最后一道人工屏障,其对放射性核素的阻滞性能将直接影响到处置库的长期稳定性和安全性。以具有低渗透性和良好的膨胀自愈性的膨润土作为集成缓冲材料的基材,以沸石和黄铁矿作为矿物添加剂,三者按照质量比为63∶27∶10均匀混合构成集成缓冲材料B7ZP,并采用恒定源扩散实验分析了锶在干密度为1.70g/cm~3试样中的扩散特性,结果表明,B7ZP缓冲材料对锶具有良好的阻滞性能,其表观扩散系数为3.30×10~(-12) m~2/s。同时,以多孔介质污染物迁移理论为依据,建立了锶在集成缓冲材料B7ZP中迁移的对流-弥散-吸附多场耦合方程,并应用Matlab软件分析了不同的时间尺度、渗流速率、表观扩散系数和阻滞因子等因素下集成缓冲材料B7ZP对锶的长期阻滞性能,为高放废物处置库的缓冲材料设计和长期阻滞性能评价提供科学依据。