模拟高放玻璃固化体在低氧地下水中的长期蚀变行为研究

高放废物玻璃固化体的长期蚀变行为对深地质处置场的安全评价非常重要。玻璃蚀变在正常条件下发展缓慢,为了能够预测玻璃的长期蚀变行为,本研究采用了150℃下,玻璃体表面积与浸泡溶液体积比(S/V)为6 000 m-1的粉末静态浸泡法(PCT法)来加快腐蚀进度,用扫描电子显微镜-X射线能谱仪(SEM-EDS)和X射线衍射仪(XRD)分析了固体样品表面形貌和二次矿物相,用电感耦合等离子体发射光谱仪(ICP-AES)分析了浸出液中的元素含量。结果表明,模拟高放废物玻璃体在遭受苦咸地下水长期浸泡的后果是表面生成蜂窝状富Mg和Fe的页硅酸盐和铝硅酸盐矿相,这些二次矿相主要是绿脱石[Na0.3Fe2Si4O1(0OH)2.4H2O]、蒙脱石[Ca0.(2Al,Mg)2Si4O1(0OH)2.4H2O]、发光沸石([Na2,K2,Ca)Al2Si10O24.7H2O]和斜发沸石([Na,K,Ca)5Al6Si30O7.218H2O]等矿物。玻璃的溶解进一步加深后,B和Na会以硼砂形式浸出。页硅酸盐矿物的形成会加快玻璃的溶解速度,重新恢复的最大速率要比之前稳定的速率高出约4倍。     

地下水穿透情形下玻璃固化体的蚀变研究

玻璃固化体在高放废物处置库中的长期处置行为，是处置库安全评价的关键环节之一。本研究模拟极端情形下，地下水穿透包装容器与固化体接触后，固化体中元素的浸出和蚀变行为。结果表明，地下水与固化体接触后，各元素的浸出浓度迅速增大，在200 d后逐渐下降并趋于稳定；温度对固化体中不同元素浸出速率的影响不同，B和Si的浸出速率随温度的增加而增大，U和Re的浸出速率随温度的降低而增大；固化体蚀变程度随温度的升高而加重，但其蚀变层的形成会阻滞元素在其中的扩散，客观上降低了固化体蚀变速率；富Si处置环境有利于抑制固化体中元素的浸出。     

重离子和γ辐照硼硅酸盐玻璃的辐照效应对比

为研究玻璃固化体在不同辐照方式下的宏观和微观结果及其对应关系,使用重离子和γ辐照硼硅酸盐玻璃,分别测量了辐照后硼硅酸盐玻璃的硬度和模量变化。发现在γ辐照条件下,直到吸收剂量达到6×10~6 Gy,硼硅酸盐玻璃的宏观性质(硬度和模量)均未发生明显改变。在Xe离子辐照条件下,当辐照剂量达到0.1 dpa时,硬度和模量减少达饱和值。此外,测量了γ辐照后的硼硅酸盐玻璃的吸收光谱,得到了辐照后硼硅酸盐玻璃带隙随吸收剂量的变化规律,讨论了辐照产生的微观缺陷来源和产生机理。发现重离子辐照产生的硬度和模量的下降主要来源于玻璃网络结构的断裂,而重离子的核能量沉积是造成网络体结构断裂的主要原因。结合γ辐照样品的吸收光谱结果,通过对比γ射线和重离子辐照后的样品硬度和模量变化不同趋势可发现:γ辐照会在硼硅酸盐玻璃中产生微观缺陷(非桥氧空位色心等),这些缺陷主要来源于网络体末端与钠相连的键的断裂。而网络体末端的断裂不影响硼硅酸盐玻璃的网络体结构,所以γ辐照产生的缺陷不会引起硼硅酸盐玻璃的硬度和模量变化。     

重离子和γ辐照硼硅酸盐玻璃的辐照效应对比

为研究玻璃固化体在不同辐照方式下的宏观和微观结果及其对应关系，使用重离子和γ辐照硼硅酸盐玻璃，分别测量了辐照后硼硅酸盐玻璃的硬度和模量变化。发现在γ辐照条件下，直到吸收剂量达到6×10⁶ Gy，硼硅酸盐玻璃的宏观性质（硬度和模量）均未发生明显改变。在Xe离子辐照条件下，当辐照剂量达到0.1 dpa时，硬度和模量减少达饱和值。此外，测量了γ辐照后的硼硅酸盐玻璃的吸收光谱，得到了辐照后硼硅酸盐玻璃带隙随吸收剂量的变化规律，讨论了辐照产生的微观缺陷来源和产生机理。发现重离子辐照产生的硬度和模量的下降主要来源于玻璃网络结构的断裂，而重离子的核能量沉积是造成网络体结构断裂的主要原因。结合γ辐照样品的吸收光谱结果，通过对比γ射线和重离子辐照后的样品硬度和模量变化不同趋势可发现：γ辐照会在硼硅酸盐玻璃中产生微观缺陷（非桥氧空位色心等），这些缺陷主要来源于网络体末端与钠相连的键的断裂。而网络体末端的断裂不影响硼硅酸盐玻璃的网络体结构，所以γ辐照产生的缺陷不会引起硼硅酸盐玻璃的硬度和模量变化。     

模拟高放玻璃体在高温湿气中的蚀变行为

模拟地下水穿透地质处置多重屏障情形下，地下水会以高温蒸气的形式与高放玻璃体发生反应，加速玻璃体的蚀变。采用纯水和特定的盐溶液作为浸泡剂以控制体系的相对湿度分别在100%和90%左右。模拟高放玻璃体样片用聚四氟乙烯线悬挂在浸泡剂上方并在低氧箱中平衡，分别在120 ℃和150 ℃的干燥箱内反应。通过测试浸泡液中各元素的浓度和分析玻璃体表面的腐蚀产物，以表征玻璃的蚀变速率。实验发现，温度对玻璃腐蚀的影响最为明显，湿度的影响较弱。玻璃体中大多数元素在150 ℃下的浸出速率高于120 ℃下的浸出速率2～3个数量级。90 d时，玻璃体在150 ℃下已观察到显著的腐蚀。900 d后，150 ℃下玻璃体已基本完全腐蚀；而在120 ℃下的玻璃腐蚀程度很小，900 d时腐蚀厚度仍小于50 μm。150 ℃时随着腐蚀的发生，玻璃表面的SiO2腐蚀层逐渐形成以钙的硅酸盐为主的二次矿相。     

模拟高放玻璃体在高温湿气中的蚀变行为

模拟地下水穿透地质处置多重屏障情形下，地下水会以高温蒸气的形式与高放玻璃体发生反应，加速玻璃体的蚀变。采用纯水和特定的盐溶液作为浸泡剂以控制体系的相对湿度分别在100%和90%左右。模拟高放玻璃体样片用聚四氟乙烯线悬挂在浸泡剂上方并在低氧箱中平衡，分别在120 ℃和150 ℃的干燥箱内反应。通过测试浸泡液中各元素的浓度和分析玻璃体表面的腐蚀产物，以表征玻璃的蚀变速率。实验发现，温度对玻璃腐蚀的影响最为明显，湿度的影响较弱。玻璃体中大多数元素在150 ℃下的浸出速率高于120 ℃下的浸出速率2～3个数量级。90 d时，玻璃体在150 ℃下已观察到显著的腐蚀。900 d后，150 ℃下玻璃体已基本完全腐蚀；而在120 ℃下的玻璃腐蚀程度很小，900 d时腐蚀厚度仍小于50 μm。150 ℃时随着腐蚀的发生，玻璃表面的SiO2腐蚀层逐渐形成以钙的硅酸盐为主的二次矿相。     

反向气相色谱法对模拟高放玻璃固化体的表面化学性能研究

高放玻璃固化体是高放废物深地质处置的核心屏障，它们在地下水浸泡下的蚀变行为是高放废物深地质处置的研究重点之一，因此，表征固体表面物理化学参数十分重要。目前，表征固体表面特征的技术主要是扫描电镜和透射电镜测量技术，这些电镜技术能够直观地给出固体表面的形貌、     

高硅玻璃陶瓷对模拟高放废液中锕系核素的固化性能测试

为了提高玻璃固化体性能,将玻璃质的Si O2在1 200℃条件下烧结制备成高硅玻璃陶瓷固化体。测试结果表明:高硅玻璃陶瓷固化体的密度、包容量及浸出率等性能均优于硼硅酸盐玻璃固化体。     

高放废物玻璃固化过程中玻璃组成对硫酸盐溶解度的影响

硼硅酸盐玻璃固化的高放废物固化体能进行长期安全存储是已为国际所公认。然而，对于含有较高浓度硫酸根的高放废液，熔制过程中会产生分离黄色第二相（简称黄相），这是一种易溶于水的结晶物质。分析表明，玻璃固化体黄相含有碱金属和碱土金属的硫酸盐、铬酸盐和钼酸盐，并有一定量的铯、锶等裂片元素。玻璃固化体在深地质处置后，一旦受到地下水侵蚀，这些核素易浸泡出来，进入生物圈，因此，它严重危害玻璃固化体包容和隔离核素的作用，这是必须克服和避免的。     

模拟高放废物玻璃固化体在处置条件下的浸出行为研究（Ⅰ）

用模拟高放废物硼硅酸盐玻璃固化体和介质（包括膨润土、凝灰岩、沸石、氧化铁粉、去离子水和模拟地下水）构成模拟处置条件下的９个浸泡体系，研究了在有介质存在条件下，玻璃固化体浸泡后的失重，玻璃体的元素浸出和浸出液的ｐＨ值变化；研究了温度和ｐＨ对浸出的影响，求出了玻璃、水反应的表观活化能为７３．０ＫＪ／ｍｏｌ。对高放废物处置库的回填材料的选择提供了优选方案。     

Electron beam irradiation effects in Trombay nuclear waste glass

Spectroscopic investigations were carried out on electron beam irradiated sodium barium borosilicate glasses, which is the base glass for immobilization of nuclear high level radioactive waste, generated from the research reactors at Bhabha Atomic Research Centre, Trombay. This was done in order to access the defects generated in it under long term irradiation. Electron paramagnetic resonance was used to identify the defect centers generated in the borosilicate glass after irradiation. In addition, positron annihilation spectroscopy and infrared investigations were done on the samples to evaluate the radiation induced changes in the glass. It was found that, boron-oxygen and silicon based hole centers along with E′ centers are getting formed in the glass after irradiation due to the breaking of the SiO bonds at regular tetrahedron sites of SiOSi. The positron annihilation spectroscopy data gave an idea regarding the free volume size and fraction of the glasses before and after irradiation. It was seen that, after irradiation the free volume size in the glass increased with creation of additional sites. Microwave power variation and temperature variation studies suggested the formation of at least five different radicals in the irradiated glasses. The spin Hamiltonian parameter of all the radical species were determined by computer simulation. An electron paramagnetic resonance spin counting technique was employed to evaluate the defect concentration in the glasses after irradiation.     

Analysis of molecular oxygen formation in irradiated glasses: a Raman depth profile study

Depth profiles experiments have been performed by Raman spectroscopy on three alkali (Na, Li, K) borosilicate glasses irradiated with 1.8 MeV electrons at 1 and 3 GGy. These experiments show that molecular oxygen produced under β irradiation is concentrated near the glass surface according to a depth depending on the irradiation dose. Moreover, we observed that the polymerisation increase is the same in the entire volume sample. The average Si–O–Si angle decrease under irradiation is also homogeneous in the whole irradiated glass volume. From all results, we demonstrate that oxygen migrates up to the glass surface during irradiation without strong interaction with the glass network. Migration of oxygen and probably alkalis takes place through percolation channels with a possible departure of oxygen in some cases.     

电子辐照对高庙子钙基膨润土的性能影响

为研究电子辐照对高庙子钙基膨润土性能的影响,用电子辐照至不同累积受照剂量后,分析了微观结构和某些物化性质。结果表明:电子辐射作用引起了膨润土中蒙脱石晶胞参数的改变和晶粒尺寸的减小,结构中部分Si—O键和Al—OH键受到破坏,导致了阳离子交换能力和层电荷的降低、膨胀指数的减小、Si和Al溶解性增加以及对¹³⁷Cs和⁹⁹Tc吸附能力的减弱;与未经辐照样品相比,累积辐照至5 MGy时,样品的膨胀指数减小了18.65%,¹³⁷Cs和⁹⁹Tc分配系数分别减小了50.85%和58.58%,Si和Al溶解性分别增加了125%和199%。在地质处置回填材料设计和安全评价中,必须考虑电子辐射作用对膨润土性能的影响。     

Physical properties and leaching behaviour of spent fuel BISO coated particles

The safe disposal in a geological repository is proposed for the spent fuel elements obtained from operation of High Temperature Reactors. The behavior of the fuel particles under disposal conditions is a key question for the long-term nuclear waste disposal. In the present work, the spent fuel BISO coated particles, which have been irradiated to a burn-up of 10% FIMA, were studied. The size and morphological characteristics of the coated particles were investigated by the using of optical and SEM microscopy. The distribution of the ¹³⁷Cs amount in the coated particle was studied in detail. It was shown the activity was concentrated mainly inside the kernels and in the carbon buffer layer, while the outside carbon layer contained 0.1% of the total ¹³⁷Cs only. Further, the thoria-based (Th_0.834U_0.166)O₂ kernels were mechanically isolated from the coated particles, and their solution behavior was studied using the flow through experiments. In all experiments the average flow rate was ∼7–8 ml/day. Dissolution of irradiated and unirradiated kernels in HCl solution with the different value of pH (from 0 to 5) was investigated at the temperatures 90, 55 and 20 °C. The amounts of the radionuclide leached in solutions were determined by ACP-MS, γ- und α-spectrometry. On the basis of the obtained results the important leaching characteristics such as the normalized leaching rate, the activation energy value for the release of the different radionuclides were calculated.     

Effects of α-radiation on a direct disposal system for spent nuclear fuel – (2) review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

The Japanese geological disposal programme has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter “direct disposal of SF”) as an alternative management option other than reprocessing followed by vitrification and deep geological disposal of high-level radioactive waste (HLW). In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of α-radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.     

Immobilization of inert TRISO-coated fuel in glass for geological disposal

Vitrification of TRISO-coated gas reactor fuel particles was achieved via two methods: glass melting and sintering. Inert TRISO-coated fuel particles and a borosilicate glass were used. With glass melting at 1200-1300 °C floatation and decomposition of carbon and silicon carbide occurred. Thermal pre-treatment of the particles for oxidation of pyrocarbon did not improve the coating properties of the glass. During cooling most of the particles floated and sorbed on the crucible or mold walls. The sintered glass at 700 °C showed better coating properties of the TRISO-coated fuel particles despite higher porosity compared to glass made by melting. Aqueous leaching properties of glass with particles are similar regardless the mode of fabrication, indicating the good chemical durability of the sintered glass. Sintered glasses may constitute a good technique for TRISO-coated fuel particles immobilization for an eventual deep geological disposal.     

Re(Ⅶ)在辐照膨润土中的扩散行为研究

高庙子(GMZ)膨润土在长期深地质处置过程中会受到高放废物的辐照,物理化学性质可能会发生改变。本文采用贯穿扩散实验法,研究了pH=3.0和10.0条件下,Re(Ⅶ)在1MGy辐照的蒙脱石和膨润土中的扩散行为,并与未辐照的蒙脱石和膨润土相比较,得到了有效扩散系数De和有效孔隙率εacc。结果表明:pH=3.0时,De=(1.00~1.49)×10~(-11) m~2/s,εacc=0.09~0.14;pH=10.0时,De=(1.50~2.10)×10~(-11) m~2/s,εacc=0.07~0.11。随着酸度的增加,有效扩散系数和孔隙率基本不变,说明pH值对Re(Ⅶ)的扩散过程无影响。在蒙脱石中,辐照后的有效扩散系数降低;而在膨润土中,辐照后的有效扩散系数升高,这可能是由于γ辐照改变了膨润土除蒙脱石以外黏土的微观结构,有效孔隙率的增加导致Re(Ⅶ)扩散通道的宽度增加。但由于有效扩散系数和有效孔隙率的改变不大,γ辐照对Re(Ⅶ)的表观扩散系数的影响不明显。此外,De与εacc的关系可通过Archie定律表述,胶结因子n=1.7~2.4。     

Radiation damage of alkali borate glasses for application in safe nuclear waste disposal

The Electron Paramagnetic Resonance technique has been used to study the time decay of paramagnetic species induced by gamma irradiation and the radiation hardness of different alkali borate glasses for their application in safe nuclear waste disposal. Glasses with different composition have been prepared by conventional melt-quenching. Glass compositions have been chosen to elucidate the role of different alkali cations and of aluminium oxide on the borate glass network. The paramagnetic states detected in these glasses have been attributed, according to the literature, to the formation of hole centers associated with threefold coordinated boron. The results indicate that the time decay trend of the different glasses is slow and that the constant decay does not appear related to the chemical composition. Moreover, the undesired strong fading of the radiation-induced signal during the first 24 h after irradiation, observable in silicate glasses has not been detected. Although no species detectable by a X band spectrometer have been generated, the interaction of lithium borate glasses with air seem to accelerate the system decay rate. Annealing was finally performed and optimized, investigating the correlation between the chemical composition and the radiation damage recovery.     

Identification of radiation treatment of wheat （Triticum aestivum. L）and rice （Oryza sativa. L） samples using thermoluminescence of contaminating minerals

Food irradiation is gaining popularity worldwide and this technology is important to improve quality and reduce the post harvest losses of food. Because of the rapid commercialization of irradiated foods throughout the world, compliance of different regulations relating to use of technology in different countries and demand of consumers for clear labelling of irradiated foods, there is need for the development of analytical methods to detect radiation treatment of food. Among several methods studied so far, thermoluminescence （TL） is an important method that can be used to find out the irradiation history of food that contain even a very minute amount of dust particles. In this study, the irradiated and unirradiated wheat and rice samples were analyzed using the TL method. The samples were purchased from the local market of Peshawar and irradiated to radiation doses of 0.5 and 1.0 kGy using Co-60 gamma irradiator at the Nuclear Institute for Food and Agriculture （NIFA）, Peshawar. The mineral contaminants were isolated by jet water, ultrasonic treatment, and density gradient. TL glow curves of the isolated minerals from irradiated and unirradiated samples were recorded between the temperature ranges of 50-500℃ using a TL reader. Generally, the glow curves for irradiated samples showed much higher TL intensities （TL1） than the unirradiated samples. The results were normalized by rerradiation of mineral samples to gamma-ray dose of 1.0 kGy followed by determination of the second glow curves （TL2）. The ratio of the area of first glow curve to that of second glow curve （TL1/TL2） was calculated for selected temperature intervals and compared with the recommended values for unirradiated and irradiated samples. Finally, the shapes of the glow curves for irradiated and unirradiated samples were also analyzed. On the basis of these results （comparison of TL-intensities, TL1/TL2 ratios and shapes of the glow curves）, all the irradiated and unirradiated samples of wheat and rice were unequivocally identified.