On the pH dependence of leaching of nuclear waste glasses

Two simulated nuclear waste glasses were leached for periods up to one month at 90°C in high-purity water, following standard MCC-1 test procedures. The changes in composition of the surface layers were determined using ESCA and SIMS, which analyses layers of different depths. The results are discussed with reference to the different pH values in the tests performed.     

The ion exchange phase in corrosion of nuclear waste glasses

The diffusion-controlled ion exchange phase in the corrosion of nuclear waste borosilicate glasses has been examined using the Doremus’ model accounting for interdiffusion and exchange of the cations in the glass with protons from the water. Ion exchange is the principal radionuclide release mechanism for conditions when glass network hydrolysis is suppressed, such as in silica-saturated solutions when ion exchange may persist over geological time scales. In dilute aqueous solutions ion exchange controls the initial cation release and can dominate for tens and many hundreds of years if temperatures are low at low and neutral pH. Ion exchange rates are shown to have inverse square root time dependences, an Arrhenius-controlled temperature relationship and a 10^−0.5pH dependence with the pH of the contacting water. Due to radioactive decay the radionuclide releases from nuclear waste glasses are limited to certain upper values, which can be calculated based on available experimental data.     

Vitrification of high chrome oxide nuclear waste in iron phosphate glasses

A simulated high level waste (HLW) containing 4 mass% chrome oxide, whose overall composition is representative of the high chrome oxide wastes at Hanford WA USA, was easily vitrified in a phosphate glass at temperatures ranging from 1150 °C, for waste loadings of 55 mass%, to 1250 °C for waste loadings of 75 mass%. Even at these high waste loadings, these wasteforms had an excellent chemical durability. The best chemical durability was achieved when the O/(Si + P) atomic ratio was between 3.5 and 3.8. These wasteforms were also resistant to crystallization although trace amounts of crystalline Cr₂O₃ were present in wasteforms containing more than 70 mass% HLW. It is concluded that up to 45 mass% of the total HLW at Hanford, especially that containing as high as 4.5 mass% chrome oxide, could be directly vitrified into an iron phosphate glass, that meets all of the current chemical durability requirements by simply adding 25-35 mass% P₂O₅ to the waste and melting the mixture at 1150-1250 °C for a few (<6) hours.     

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.     

Solubility of actinide surrogates in nuclear glasses

This paper discusses the results of a study of actinide surrogates in a nuclear borosilicate glass to understand the effect of processing conditions (temperature and oxidizing versus reducing conditions) on the solubility limits of these elements. The incorporation of cerium oxide, hafnium oxide, and neodymium oxide in this borosilicate glass was investigated. Cerium is a possible surrogate for tetravalent and trivalent actinides, hafnium for tetravalent actinides, and neodymium for trivalent actinides. The material homogeneity was studied by optical, scanning electron microscopy. Cerium L_III XANES spectroscopy showed that the Ce³⁺/Ce_total ratio increased from about 0.5 to 0.9 as the processing temperature increased from 1100 to 1400 °C. Cerium L_III XANES spectroscopy also confirmed that the increased Ce solubility in glasses melted under reducing conditions was due to complete reduction of all the cerium in the glass. The most significant results pointed out in the current study are that the solubility limits of the actinide surrogates increases with the processing temperature and that Ce³⁺ is shown to be more soluble than Ce⁴⁺ in this borosilicate glass.     

Experience in SIMS surface analysis of radioactive waste glasses

In order to predict the long term corrosion behaviour of high level radioactive vitrified waste under repository conditions, surface analytical tools play an important role. The paper describes the experience gained with a shielded secondary ion mass spectrometer in active glass surface analysis. To improve on the qualitative character of in depth profile data, relative sensitivity factors are applied and an internal elemental standard of known surface concentration is used to quantify the profile concentration results.     

核废料核素价值研究

迄止21世纪初，全世界30多个国家和地区的450多座核电站在运行，为全球提供的电力超过总电力的16%。我国核电起步较晚，但到2010年，投入运行的核电机组也将超过20个。全世界已运行的核电机组绝大多数为轻水堆。因此，轻水堆电站，特别是压水堆(PWR)核电站发展中提出的问题将在很大程度上左右裂变核能的持续发展。经过多年的研究和发展，商业规模的压水堆核电站已可安全可靠、经济高效地运行。     

Corrosion issues in nuclear waste disposal

This paper summarized some corrosion issues specific to nuclear waste disposal and illustrates them by the French geological clay concept for the reliable prediction of container degradation rate and engineering barrier integrity over extended periods, up to several thousands years. Among the items, the following are included:

• The importance of the underground repository conditions.

• The necessity of developing comprehensive semi-empirical models and also predictive models that must be based on the mechanisms of corrosion phenomena.

• The use of archaeological artefacts to demonstrate the feasibility of long term storage and to provide a database for testing and validating the models.

桶装核废物的非破坏性分析(续)

有源γ射线(X射线)法可进一步分为吸收谱(吸收测量)法和有源诱发射线法。吸收谱法的分析原理是：当一定量的射线透射样品时,样品内的各元素具有正好吸收某一能量的射线的特性,从而使透射的射线强度大大减弱,这种减弱的多少与该元素的含量有关,这一现象称为     

桶装核废物的非破坏性分析

综述了桶装核废物非破坏性分析常规方法(分段γ扫描法、无源中子法、有源中子法、有源γ射线法)及基本原理、研究与进展的最新概况，介绍了生产相关仪器和设备的主要厂商。     

Processing glass-pyrochlore composites for nuclear waste encapsulation

Glass matrix composites have been developed as alternative materials to immobilize nuclear solid waste, in particular actinides. These composites are made of soda borosilicate glass matrix, into which particles of lanthanum zirconate pyrochlore are encapsulated in concentrations of 30 vol.%. The fabrication process involves powder mixing followed by hot-pressing. At the relatively low processing temperature used (620 °C), the pyrochlore crystalline structure of the zirconate, which is relevant for containment of radioactive nuclei, remains unaltered. The microstructure of the composites exhibits a homogeneous distribution of isolated pyrochlore particles in the glass matrix and strong bonding at the matrix-particle interfaces. Hot-pressing was found to lead to high densification (95% th.d.) of the composite. The materials are characterized by relatively high elastic modulus, flexural strength, hardness and fracture toughness. A numerical approach using a microstructure-based finite element solver was used in order to investigate the mechanical properties of the composites.     

Transient thermal response in nuclear waste repositories

In this paper, it is shown that a previously reported non-linear, one-dimensional, theoretical approximation simplifies — from a computational point of view — the calculation of the time-decay temperature field in nuclear waste repositories (NWR). This conclusion has been reached after solving, by using the control volume numerical method, the full three dimensional, transient, non-linear heat diffusion equation. The transient thermal field in a rock salt repository, is analytically solved and numerically predicted, along 100 years, after the disposal of a high-level waste (HLW). The nuclear waste, with a half-life of 32.9 years, releases an exponentially time dependent heat flux with 12 W m⁻² as the initial thermal load. Two cases are studied, in the first one it is assumed that the conductivity (k) and the volumetric heat capacity ρc_p of the host rock (diffusion domain) remain constant (linear case), whereas in the second one, a more realistic situation is analysed. In this last case, the conductivity of the rock salt varies as a function of the temperature field and the product ρ×c_p remains constant (non-linear case). In order to observe the effect of the salt conductivity (constant or variable) on the repository temperature distribution, a comparison of both cases is performed. It is concluded, that the theoretical model, which provides an analytical solution of the thermal fields may be a powerful low cost method for design purposes.     

Deep-Burn: making nuclear waste transmutation practical

In the Deep-Burn concept, destruction of the transuranic component of light water reactor (LWR) waste is carried out in one burn-up cycle, accomplishing the virtually complete destruction of weapons-usable materials (Plutonium-239), and up to 90% of all transuranic waste, including the near totality of Neptunium-237 (the most mobile actinide in the repository environment) and its precursor, Americium-241.Waste destruction would be performed rapidly, without multiple reprocessing of plutonium, thus eliminating the risks of repeated handling of weapons-usable material and limiting the generation of secondary waste. There appears to be no incentive in continuing the destruction of waste beyond this level.An essential feature of the Deep-Burn Transmuter is the use of ceramic-coated fuel particles that provide very strong containment and are highly resistant to irradiation, thereby allowing very extensive destruction levels (“Deep Burn”) in the one pass, using gas-cooled modular helium reactor (MHR) technology developed for high-efficiency energy production. The fixed moderator (graphite) and neutronically transparent coolant (helium) provide a unique neutron energy spectrum to cause Deep-Burn, and inherent safety features, specific to the destruction of nuclear waste, that are not found in any other design.Deep-Burn technology could be available for deployment in a relatively short time, thus contributing effectively to waste problem solutions. Extensive modeling effort has led to conceptual Deep-Burn designs that can achieve high waste destruction levels (70% in critical mode, 90% in with a supplemental subcritical step) within the operational envelope of commercial MHR operation, including long refueling intervals and the highly efficient production of energy (approximately 50%). To the plant operator, a Deep-Burn Transmuter will be identical to its commercial reactor counterpart.     

Vitrification of nuclear waste with low heat generation

Analyses of thermal processes in the glass melter and storage container are given for vitrification of defense waste.     

The efficient transport and storage of nuclear waste

The arrangements that BNFL has for managing the return and storage of highly active waste generated by its reprocessing operations at Sellafield are described. The philosophy of substitution of low and intermediate wastes for an increase in high level waste is discussed. Also the concepts of flask design so that the vitrified residue waste storage is optimized are described.     

Three-dimensional thermal analysis of in-floor type nuclear waste repository for a ceramic waste form

A thermal model is constructed and analyses are performed for an ‘in-floor’ type nuclear waste repository in granitic rock for a high level nuclear waste (HLW)-bearing ceramic waste form (synroc). Transient calculations for a three-dimensional (3-D) model have been carried out for both 20 and 10 wt.% HLW-bearing synroc, for surface cooling periods between reactor discharge and geological disposal varying from 5 to 40 years. This study investigates the temperature distribution in one of the boreholes of a hypothetical tunnel for a basic geometrical setting as well as the effect of varying the distance between adjacent boreholes and the distance between adjacent tunnels. The temperatures in the repository were found to be sensitive to the interim surface cooling period as well as the amount of waste loaded. The results showed that decreasing the spacing between the canisters has a more pronounced effect on the temperature field than decreasing the spacing between the tunnels.     

Structural role of molybdenum in nuclear glasses: an EXAFS study

The Mo environment has been investigated in inactive nuclear glasses using extended X-ray absorption spectroscopy (XAS). Mo is present in a tetrahedron coordinated to oxygen in the form of molybdate groups [MoO₄]²⁻ (d(Mo-O)=1.78 Å). This surrounding is not affected by the presence of noble metal phases in the nuclear glass. Relying on the XAS results, on the bond-valence model and on molecular dynamics simulations of a simplified borosilicate model glass, we show that these groups are not directly linked to the borosilicate network but rather located within alkali and alkaline-earth rich domains in the glass. This specific location in the glass network is a way to understand the low solubility of Mo in glasses melted under oxidizing conditions. It also explains the possible phase separation of a yellow phase enriched in alkali molybdates in molten nuclear glasses or the nucleation of calcium molybdates during thermal aging of these glasses. Boron coordination changes in the molten and the glassy states may explain the difference in the composition of the crystalline molybdates, as they exert a direct influence on the activity of alkalis in borosilicate glasses and melts.     

Mechanical properties of uranium

The paper describes the results of mechanical tests on uranium at room temperature and at elevated temperatures. Data are given on the hardness of uranium in the temperature range 20–600 C, flow pressure on extrusion in the ^– and ^– phase ranges, tensile properties and impact strength at temperatures of the ^–, ^–, and ^– phases. The anisotropic behavior of the individual grains of coarsegrained uranium during mechanical tests has been elucidated. It is shown that the existence of allotropic transformations and the difference in the crystal structure of the modifications of uranium influence its mechanical properties to a marked degree. It is also shown that the mechanical properties depend upon the carbon content of the uranium.     

Cu-Zr-Al-Sm非晶合金的γ射线屏蔽性能研究

采用铜模吸铸法制备了(Cu50Zr45Al5)100-xSmx(x=0,1,2,3,4)板状合金试样。用差热扫描量热分析(Differential scanning calorimetry,DSC)、差热分析(Differential thermal analysis,DTA)和X射线衍射分析(X-ray diffraction,XRD)研究了Sm微合金化对Cu50Zr45Al5非晶合金结构与热稳定性的影响。采用Na I(T1)单晶γ闪烁能谱仪测定了(Cu50Zr45Al5)99Sm1非晶合金对γ射线的线性吸收系数。结果表明,当Sm含量低于3%(at%)时制备的板状合金为完全的非晶结构,超过3%会有Cu10Zr7和Zr Cu等金属间化合物析出。Sm的微量添加增加了合金的热稳定性。制备的(Cu50Zr45Al5)99Sm1非晶合金对137Cs和60Co两种γ射线的线性吸收系数分别为0.49 cm-1和0.69 cm-1。其屏蔽性能介于Pb和Al之间,具有潜在的应用价值。