ZnO和CaO对模拟高放废液硅酸盐玻璃固化体性能的影响研究

针对有些高放废液含有较多Fe、Cr、Ni过渡金属元素,在玻璃固化工艺过程中易于形成晶体,导致熔融玻璃体的黏度增加、化学稳定性变差以及工艺过程中易出现出料口堵塞等问题,研究了废物包容量为15%和20%、添加ZnO(5.6%)和CaO(1.75%)的配方对形成的4种玻璃固化体的物理性能(密度、硬度、断裂韧性)、化学性能(产品一致性测试和蒸汽腐蚀测试)和结构(X射线衍射析晶分析、拉曼光谱分析)的影响。研究分析显示,提高废物包容量至20%以及添加ZnO和CaO均可促进硼硅酸盐玻璃固化体网络结构的稳定性和化学稳定性,并增强玻璃体的密度,提高硬度;但玻璃固化体的高温黏度升高,断裂韧性下降。     

Vitrification and testing of a Hanford high-level waste sample. Part 1: Glass fabrication, and chemical and radiochemical analysis

The Hanford radioactive tank waste will be separated into low-activity waste and high-level waste that will both be vitrified into borosilicate glasses. To demonstrate the feasibility of vitrification and the durability of the high-level waste glass, a high-level waste sample from Tank AZ-101 was processed to glass, analyzed with respect to chemical composition, radionuclide content, waste loading, and the presence of crystalline phases and then tested for leachability. The glass was analyzed with inductively coupled plasma-atomic emission spectroscopy, inductively coupled plasma-mass spectrometry, γ-energy spectrometry, α-spectrometry, and liquid scintillation counting. The WISE Uranium Project calculator was used to calculate the main sources of radioactivity to the year 3115. The observed crystallinity and the results of leachability testing of the glass will be reported in Part 2 of this paper.     

添加剂对模拟含氟放射性废液水泥固化性能的影响

采用沸石、蛭石、硅灰、石英砂作为添加剂对模拟含氟放射性废液(主要含有Cs+、Sr2+、F-等)水泥固化性能的影响进行了研究。通过添加不同的添加剂,测定水泥浆的初终凝时间、流动度和温升;测定固化体养护28d后的抗压强度及浸泡、冻融试验后的抗压强度,并进行了抗冲击试验和模拟核素浸出试验。通过对比,得到了满足模拟含氟放射性废液水泥固化的配方,该配方制成的水泥固化试块性能达到了GB 14569.1—2011的要求。     

模拟含锶废物铁磷酸盐玻璃固化体的化学稳定性

针对我国高放废液全分离流程中产出的锶废物组成特点,设计了用铁磷酸盐玻璃固化锶废物的配方。用红外光谱（IR）研究了玻璃固化体的结构,用Product Consistency Test(PCT）试验方法研究了玻璃固化体的化学稳定性。研究表明,在所选的配方组成范围内,所熔制的玻璃固化体均有较好的化学稳定性。当配料中模拟含锶废物的含量为24～28%(wt)、FeO3的含量大于24%(wt)、O/P（氧磷摩尔比）为3.5～3.6时,玻璃固化体的化学稳定性最好。     

模拟放射性含氟废液水泥固化配方的研究

本文以沸石、硅灰、石英砂为添加剂,按照质量比m（沸石）∶m(硅灰)∶m（石英砂）∶m（水泥）=1∶1∶3∶10配方对模拟放射性含氟废液进行水泥固化。由配方得到的水泥浆流动度和初、终凝时间满足桶内固化要求。测定了水泥固化体28 d的抗压强度、抗浸泡性和抗冻融性实验后的强度损失,进行了抗冲击性能测试和模拟核素浸出实验。结果表明,该配方可有效地固化模拟放射性含氟废液,固化体28 d抗压强度、各项实验强度损失和模拟核素浸出率均满足GB 14569.1-2011的要求。水泥固化体的F^-浸出率很低,XRD显示F^-以CaF₂形式存在。废液中F^-质量分数控制在1%较为合适,此时水泥固化体终凝时间为14 h,F^-的42 d浸出率为2.54×10^-3 cm/d。     

Diffusion of Neptunium in Simulated High Level Waste Glass

Neptunium in high level radioactive wastes has to be retained in glasses through geological period from the point of biological toxicity for more than million years. Neptunium-237 diffusion in borosilicate glass with simulated wastes of 26.4% was investigated in the temperature range of 400–600°C by the use of α-degradation method. The energy loss rate dE/dx of α-particles, which is necessary in order to determine diffusion coefficients by the α-degradation method, was calculated for the waste glass.

The penetration depth of α-particle with 4.787MeV from ²³⁷Np was 17 μm, which gives a limit in applying the α-degradation method in the waste glass. The temperature dependence of the diffusion coefficient of Np in the waste glass was given by

D _Np=3.67 exp(-55,900/RT) (cm²/s), in which the activation energy of the diffusion was 55.9 kcal/mol. It was clarified that Np is one of the elements with the lowest mobility in waste glasses.     

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.     

Leach Rates and Thermal Properties of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate (LIP) glasses loaded with a simulated high-level nuclear waste were studied on their leach rates and thermal properties.

The obtained results showed that the phosphate glass matrix consisting of lead monoxide, phosphorus pentoxide and ferric oxide of 56:35:9w/₀ is able to vitrify the waste, pretreated with formic acid to remove Zr, to about 15 w/₀ at 950°C. The leach rate of the vitrified waste glass was in the order of 10^?7 g/cm².d at 110°C, which is low compared with that of the borosilicate glass waste form. Increasing the phosphorus pentoxide content of the matrix to higher than 35% enabled it to produce the glass form with the waste near 20 w/₀ at 950°C, but this increase rendered the glass waste form more soluble than the former. Thermal properties such as thermal expansion coefficient, critical cooling rate for vitrification and temperatures of glass transition, softening and maximum rate of crystallization were measured and discussed.

Removing Na ions from wastes improves considerably both the leach rate and the thermal stability of the LIP glass waste form.     

Angular Distribution of Gadolinium Vapor Produced by Electron Beam Heating

Static corrosion tests were performed for the glass phase of a simulated waste form of non-combustible radioactive low-level waste to study a basic aqueous corrosion behavior. The waste form, which was fabricated from simulated waste sample by use of in-can type induction-heated melting, consists of two separated phases; a glass phase and a metal phase. Tests were performed for the glass phase from two types of the waste form with different chemical composition at 35°C and S/V ratio of 2,600 m^?1. The glass phase with both types showed an incongruent dissolution similar to conventional high-level radioactive waste (HLW) glasses, i.e., the normalized elemental mass loss (NL_i) for soluble elements such as B and Na continued to increase after the saturation of insoluble elements such as Si, A1 and Ca. The NL_i for B increased in proportion to the square root of time except for early stage, which suggests that the rate of the long-term dissolution or alteration may be controlled by a diffusion process. Potential secondary phases forming as the results of incongruent dissolution were estimated to be kaolinite and calcite by comparison of the measured solution data with the thermodynamically calculated phase stability relationships. These results suggest that the glass phase has a potential chemical durability not so different from conventional HLW glasses.     

Leach Rates of Lead-Iron Phosphate Glass Waste Forms

Lead-iron phosphate glasses loaded with simulated high-level nuclear wastes at temperatures between 900 and 1,100°C were studied on their soaking behavior in distilled water by means of leachate solution analysis.

The obtained results showed that the leach rates of the glass waste forms were at least 10 to 100 times lower than that of the currently investigated borosilicate glass, even though the selective release of Na ion from the forms was observed. Zirconium of the waste led the glass to partial crystallization at 900°C, but was able to be incorporated in the glass at near 1,100°C.

The liquid chromatographic analysis of poly-phosphate ions in the leachate solution revealed that the low leachability of the glass forms was brought about by a certain degree of depoly-merization of long poly-phosphate chains of lead metaphosphate caused by the addition of ferric oxide.     

含Pu废物的玻璃和玻璃陶瓷固化基材研究进展

对于239Pu含量较高且很难回收利用的含Pu废物,在安全处置前须进行妥善的固化处理。玻璃和玻璃陶瓷因在制备方面具有较陶瓷简单的工艺、低廉的成本和高效的产出被认为是目前处理含Pu废物综合优势明显的固化基材,因而得到了广泛和深入的研究。本文对碱硼硅酸盐玻璃、镧硼硅酸盐玻璃、铁磷酸盐玻璃以及含钙钛锆石、烧绿石或独居石结晶相的玻璃陶瓷等在含Pu废物固化方面的研究进展进行了综述,包括其组分、Pu包容量和化学稳定性,并进行了对比分析,认为在对玻璃固化基材继续研究与应用的基础上,玻璃陶瓷有望成为固化绝大多数含Pu废物的较佳选择。     

The precision of product consistency tests conducted with a glass-bonded ceramic waste form

The product consistency test (PCT) that is used for qualification of borosilicate high-level radioactive waste (HLW) glasses for disposal can be used for the same purpose in the qualification of the glass-bonded sodalite ceramic waste form (CWF). The CWF was developed to immobilize radioactive salt wastes generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuels. An interlaboratory study was conducted to measure the precision of PCTs conducted with the CWF for comparison with the precision of PCTs conducted with HLW glasses. The six independent sets of triplicate PCT results generated in the study were used to calculate the intralaboratory and interlaboratory consistency based on the concentrations of Al, B, Na, and Si in the test solutions. The results indicate that PCTs can be conducted as precisely with the CWF as with HLW glasses. For example, the values of the reproducibility standard deviation for Al, B, Na, and Si were 1.36, 0.347, 3.40, and 2.97 mg/l for PCT with CWF. These values are within the range of values measured for borosilicate glasses, including reference HLW glasses.     

Solidification of simulated actinides by natural zircon

Natural zircon was used as precursor material to produce a zircon waste form bearing 20wt% simulated actinides (Nd2O3 and UO2) through a solid state reaction by a typical synroc fabrication process. The fabricated zircon waste form has relatively good physical properties (density 5.09g/cm^3, open porosity 4.0%, Vickers hardness 715kg/mm^2). The XRD, SEM/EDS and TEM/EDS analyses indicate that there are zircon phases containing waste elements formed through the reaction. The chemical durability and radiation stability are determined by the MCC-1 method and heavy ion irradiation; the results show that the zircon waste form is highly leach resistance and relatively stable under irradiation (amorphous dose 0.7dpa). From this study, the method of using a natural mineral to solidify radioactive waste has proven to be feasible.     

DIADDHEM set-up: New IBA facility for studying the helium behavior in nuclear glasses

The immobilization of fission products and minor actinides by vitrification is the reference process for industrial management of high-level radioactive wastes generated from spent fuel reprocessing. The glassy matrix is subjected to radiation damage and radiogenic helium generation due to the alpha decays of minor actinides.A specific experimental study has been conducted to better understand the behavior of helium and its diffusion mechanisms in the borosilicate glass. Helium production is simulated by external irradiation with ³He⁺ ions at a concentration (2 × 10¹⁵ He cm^?2) equivalent to the one obtained after 1000 years of glass storage. He diffusion coefficients as function of temperature are extracted from the evolution of the depth profiles after annealing. The ³He(d, α) ¹H Nuclear Reaction Analysis (NRA) technique is successfully used for in situ low-temperature measurements of depth profiles. Its high depth resolution allows detecting helium mobility at a temperature as low as 250 K and the presence of a trapped helium fraction. The good agreement of our first values of diffusion coefficients with the literature data highlights the relevance of the implantation technique in the study of helium diffusion mechanisms in borosilicate glasses.     

Use of poly(methyl methacrylate) in radioactive waste management: II. Monte Carlo simulations

Radioactive waste is generated from the nuclear applications and it should properly be managed in a radioactive waste management system. Different methods are available for treatment and conditioning of radioactive waste. Polymers can be used in the radioactive waste management as an embedding matrix. Poly(methyl methacrylate (PMMA) is a possible candidate material that can be used in the low level radioactive waste management. In this study, based on total resistible dose for PMMA, maximum waste activity that can be embedded into a waste drum was found via Monte Carlo simulations. In addition, Monte Carlo simulations for radioactive waste embedded into above mentioned polymer was performed and the dose rate distribution in the polymer matrix was determined for the initial and different periods of 15.1, 30.2 and 302 years after embedding of waste. Changes of mechanical properties in the polymer embedded waste drum was simulated for PMMA embedded waste matrices based on experimental data.     

Recrystallisation of amorphous natural brannerite through annealing: The effect of radiation damage on the chemical durability of brannerite

An amorphous natural brannerite sample was chosen to study the annealing of the radiation damage through thermal recrystallisation, and subsequently to evaluate the effect of radiation damage on its aqueous durability. Microstructural characterisation of the natural brannerite revealed minor alteration along the rim of the crystal and within cracks. The formation of UO₂ particles and soluble Pb–Ca-rich aluminosilicate glass is responsible for the higher U and Pb releases from the recrystallised brannerite. In general, natural brannerite has been shown to be resistant to dissolution over geological time, therefore minor brannerite inclusions in ceramic formulations for immobilisation of actinide-rich radioactive wastes should not have a detrimental effect on the long-term chemical durability of the wasteforms.     

离子辐照导致硼硅酸盐玻璃浸出性能的变化北大核心CSCD

在高放射性废物深地质处置过程中,放射性核素衰变引起的强辐射场会导致玻璃固化体浸出性能的变化。本工作利用15 MeV Si离子辐照模拟深地质处置强辐射场,采用MCC-1静态浸泡法在pH=9.0的KOH溶液中对硼硅酸盐玻璃分别浸泡1、3、7、14、28 d,结合电感耦合等离子体发射光谱仪、拉曼光谱仪以及扫描电镜研究了辐照前后玻璃样品的浸出行为。结果表明:随着浸出时间推移,玻璃表面会不断地腐蚀脱落;辐照后玻璃样品的浸出率明显高于未辐照玻璃样品的,蚀变层结构与元素组成无明显差异,但厚度明显增大;形成的蚀变层中Na元素和B元素基本耗尽;蚀变层中SiO_(2)网络体结构溶解,生成大量的Si—OH键及硼硅酸盐结构,并出现四配位B单元向三配位B单元的转变。     

Continuous separation of molybdenum and zirconium from simulated high-level liquid waste with a Taylor–Couette contactor

Reprocessing of spent nuclear fuels generates high-level liquid waste (HLLW) which undergoes vitrification into borosilicate glass before final geological disposal. To ensure the quality of the glass, control of the concentration of chemical species such as molybdenum (Mo), which has an adverse impact on the vitrification process, is critical. Also, zirconium (Zr) can cause crud in washing process and Zr-93 is a long-lived fission product needed to be separated. In this study, a liquid–liquid countercurrent centrifugal contactor with Taylor–Couette flow (TC contactor) was applied to practical multi-species cases. Continuous separation of Mo and Zr from a simulated HLLW with bis(2-ethylhexyl) phosphoric acid (HDEHP) as extractant has been performed. Among a variety of metals in simulated HLLW, Mo, Zr, Y, and Fe are extractable, Mo and Zr were separated from HLLW by equilibrium, and Fe/Y separation was achieved by the effect of non-equilibrium state in TC contactor. Addition of tributyl phosphate could improve extraction of Mo. This study has expanded the scope of the TC contactor to multi-species separation processes.     

放射性废物水泥固化研究进展

水泥化学的理论研究进展和新的水泥系列、混合材、外加剂及混凝土用纤维材料的研究成果,均可借鉴到放射性废物水泥固化的配方研究中。本文综述硅酸盐水泥、碱活化矿渣水泥、硫铝酸盐水泥等在放射性废物水泥固化研究中的应用现状,介绍火山灰质混合材、外加剂、纤维材料等在提高废物包容量、固化体强度、耐久性和降低核素浸出等方面的研究进展,以期为水泥固化配方的研究与开发提供新的思路。     

模拟非α低中放废液水合陶瓷固化体的水化产物和力学性能

以模拟非α低中放废液为固化处理对象,用水热法合成了非α低中放废液中和处理后沉淀的＂碱-矿渣-粉煤灰-偏高岭土＂水合陶瓷固化体。采用X射线衍射仪分析了固化体的水化产物,确定了水化产物的组成,并测试了固化体的抗压强度。研究结果表明：温度为150～180℃、废液沉淀与固化原材料的质量比值（即盐灰比）为0.10～0.30时,固化体水化产物的主要物相为方沸石,随着温度升高和反应时间延长,水化产物中方沸石的衍射峰不断增加。固化体抗压强度测试结果表明,该固化体具有较高的抗压强度,但盐灰比由0.10增加至0.30时,固化体抗压强度由26.33 MPa下降到8.46 MPa。