氧化硼对铁磷酸盐玻璃陶瓷固化体的影响

研究了不同B₂O₃掺量对铁磷酸盐玻璃陶瓷高放废物固化体结构和性能的影响。应用溶出速率法（DR）对固化体进行了化学稳定性测试,使用傅里叶变换红外光谱（FTIR）和X射线衍射（XRD）方法研究了样品的结构。研究结果表明：玻璃陶瓷固化体的主晶相为独居石;B₂O₃的引入对玻璃陶瓷固化体的化学稳定性影响较大,以10%（摩尔分数）的B₂O₃代替Fe₂O₃制得的固化体化学稳定性最佳,其28d的质量浸出率约为7.81×10^-9g•cm^-2•min^-1;试样中存在大量正磷酸基团［PO₄］^3-和少量焦磷酸基团［P₂O₇］^4-,无偏磷酸基团［PO₃］^-存在,固化体中的B主要以［BO₄］四面体基团形式存在。     

Vitrification and testing of a Hanford high-level waste sample. Part 2: Phase identification and waste form leachability

A sample of Hanford high-level radioactive waste from Tank AZ-101 was vitrified into borosilicate glass and tested to demonstrate its compliance with regulatory requirements. Compositional aspects of this study were reported in Part 1 of this paper. This second and last part presents results of crystallinity and leachability testing. Crystallinity was quantified in a glass sample heat treated according to the calculated cooling curve of glass at the centerline of a Hanford Waste Treatment Plant canister. By quantitative X-ray diffraction analysis and image analysis applied to scanning electron microscopy micrographs, the sample contained 7 mass% of spinel, a solid solution of franklinite, trevorite, and other minor spinels. Glass leachability was measured with the product consistency test and the toxicity characteristic leaching procedure. Measured data and model estimates were in reasonable agreement. Leachability results were close to those obtained for the non-radioactive simulant. Models were used to elucidate the effects of glass composition of spinel formation and to estimate effects of spinel formation on glass leachability.     

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.     

高放核废料磷酸盐玻璃陶瓷固化研究进展

磷酸盐玻璃陶瓷是固化“难溶”核废料的优异基材,具有高的废料包容量和优异的稳定性,因而,磷酸盐玻璃陶瓷固化是高放核废料固化处理的重要研究方向之一。本文简要综述了高放核废料磷酸盐玻璃陶瓷固化体的类型、固化机理、固化体设计、稳定性及其制备,并对其研究做了展望。其今后研究方向主要包括：（1） 磷酸盐玻璃固化体的中长期化学稳定性、蚀变规律和抗腐蚀机制的研究;关注其物理性能、热稳定性和辐照稳定性;（2） 磷酸盐玻璃陶瓷固化体的简洁制备工艺技术及其工艺原理,及其对设备和电极的侵蚀和寿命的影响。     

Magnesium potassium phosphate ceramic for Tc immobilization

Technetium-99, present in the US Department of Energy’s (DOE) high-level waste (HLW) as a by-product of fission reactions, poses a serious environmental threat because it has a long half-life, is highly mobile in its soluble Tc⁷⁺ oxidation state and is volatile at high temperatures. Magnesium potassium phosphate (MKP) ceramics have been developed to treat ⁹⁹Tc that has been partitioned and eluted from simulated high-level tank wastes by means of sorption processes. Waste forms were fabricated by adding MKP binder and a reducing agent (SnCl₂) to the ⁹⁹Tc-containing aqueous waste. In addition, waste forms were fabricated by first precipitating ⁹⁹Tc from the waste and subsequently solidifying it in MKP. ⁹⁹Tc loadings in the waste forms were as high as 900 ppm by weight. Waste form performance was established through various strength, leaching, and durability tests. Long-term leaching studies, as per the ANS 16.1 procedure, showed leachability indices between 11 and 14 for ⁹⁹Tc under ambient conditions. The normalized leach rate for ⁹⁹Tc, according to the product consistency test, was as low as 1.1 × 10⁻³ g/m² d. The waste forms exhibited a compressive strength of ≈30 MPa and were durable in an aqueous environment. Containment of ⁹⁹Tc in MKP ceramics is believed to be due to a combination of appropriate reducing environment (determined from Eh-pH measurements) and microencapsulation in a dense matrix.     

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。     

Characterization of monazite glass-ceramics as wasteform for simulated α-HLLW

Two monazite glass-ceramic wasteforms were sintered by mixing the lanthanum metaphosphate glass powder with the oxide powder of the components in simulated α-HLWs. The co-existence of components Al and Mo in an iron phosphate melt separated the melt into two immiscible glass melts, namely aluminum iron phosphate glass (Gb) and molybdenum iron phosphate glass (Gg). 24 wt% of ZrO₂, together with P₂O₅ and proper amounts of Fe and Mo formed a zirconium pyrophosphate glass (Gg1), which was immiscible with the phase Gg. The iron ions in the wasteforms were all in Fe³⁺, 1/3 of which was in 4-fold coordination. The O/P and O/(P + 1/3Fe³⁺) ratios for the glass phases were Gg1 3.70, Gb 3.89-3.98, Gg 4.23-4.25, and Gg1 3.58, Gb 3.47-3.42, Gg 3.74-3.69, respectively. The dissolution rates of two wasteforms were 0.3008 and 0.2598 g/m²d, respectively.     

Corrosion behavior of Inconel 690 and 693 in an iron phosphate melt

The corrosion resistance of Inconel 690 and 693 coupons submerged in an iron phosphate melt has been investigated. After 155 days in an iron phosphate melt at 1050 °C, which contained 30 wt% of a simulated low activity waste (LAW at Hanford), the weight loss of Inconel 690 and 693 was 14% and 8%, respectively. The overall corrosion rate, calculated from the initial and final dimensions of each coupon, was 1.3 and 0.7 μm/day for the Inconel 690 and 693, respectively. Scanning electron microscopy and X-ray diffraction of the submerged Inconel coupons after 155 days in the iron phosphate melt showed that an altered surface layer had formed which was depleted in nickel and consisted of a (Fe, Cr)₂O₃ solid solution. This altered layer appears to be chemically protective as indicated by the gradual reduction in weight loss which occurred with time in the iron phosphate melt. Inconel 693 appears to be a better candidate to use as an electrode in iron phosphate melts since its corrosion rate and weight loss was only about one half that of Inconel 690.     

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.     

Al2O3对独居石玻璃陶瓷固化体的影响

本文研究了Al₂O₃掺量对独居石玻璃陶瓷固化体结构和化学稳定性的影响。用傅里叶变换红外光谱（FTIR）和X射线衍射（XRD）方法表征样品结构,用溶解速率法和全谱直读等离子体发射光谱（ICP-OES）分别测定样品在浸出液中浸泡后的失重速率及各元素的浸出浓度,以研究固化体的化学稳定性。研究结果表明：当Al₂O₃掺量为4%（摩尔分数）时,在980 ℃下保温3 h得到的独居石玻璃陶瓷固化体具有较高的化学稳定性,浸泡14 d时其质量浸出率最低,约为8.1 ng/(cm²•min),其中Ce、La元素在浸出液中均未检出;固化体的主晶相为独居石,结构中含有大量稳定的正磷酸基团［PO₄］^3-和少量的焦磷酸基团［P₂O₇］^4-,不存在偏磷酸基团［PO₃］^-。     

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.     

Long-term product consistency test of simulated 90-19/Nd HLW glass

Chemical durability of 90-19/Nd glass, a simulated high-level waste (HLW) glass in contact with the groundwater was investigated with a long-term product consistency test (PCT). Generally, it is difficult to observe the long term property of HLW glass due to the slow corrosion rate in a mild condition. In order to overcome this problem, increased contacting surface (S/V = 6000 m⁻¹) and elevated temperature (150 °C) were employed to accelerate the glass corrosion evolution. The micro-morphological characteristics of the glass surface and the secondary minerals formed after the glass alteration were analyzed by SEM-EDS and XRD, and concentrations of elements in the leaching solution were determined by ICP-AES. In our experiments, two types of minerals, which have great impact on glass dissolution, were found to form on 90-19/Nd HLW glass surface when it was subjected to a long-term leaching in the groundwater. One is Mg-Fe-rich phyllosilicates with honeycomb structure; the other is aluminosilicates (zeolites). Mg and Fe in the leaching solution participated in the formation of phyllosilicates. The main components of phyllosilicates in alteration products of 90-19/Nd HLW glass are nontronite (Na_0.3Fe₂Si₄O₁₀(OH)₂·4H₂O) and montmorillonite (Ca_0.2(Al,Mg)₂Si₄O₁₀(OH)₂·4H₂O), and those of aluminosilicates are mordenite ((Na₂,K₂,Ca)Al₂Si₁₀O₂₄·7H₂O)) and clinoptilolite ((Na,K,Ca)₅Al₆Si₃₀O₇₂·18H₂O). Minerals like Ca(Mg)SO₄ and CaCO₃ with low solubility limits are prone to form precipitant on the glass surface. Appearance of the phyllosilicates and aluminosilicates result in the dissolution rate of 90-19/Nd HLW glass resumed, which is increased by several times over the stable rate. As further dissolution of the glass, both B and Na in the glass were found to leach out in borax form.     

A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

Chloride-containing radioactive wastes are generated during the pyrochemical reprocessing of Pu metal. Immobilization of these wastes in borosilicate glass or Synroc-type ceramics is not feasible due to the very low solubility of chlorides in these hosts. Alternative candidates have therefore been sought including phosphate-based glasses, crystalline ceramics and hybrid glass/ceramic systems. These studies have shown that high losses of chloride or evolution of chlorine gas from the melt make vitrification an unacceptable solution unless suitable off-gas treatment facilities capable of dealing with these corrosive by-products are available. On the other hand, both sodium aluminosilicate and calcium phosphate ceramics are capable of retaining chloride in stable mineral phases, which include sodalite, Na₈(AlSiO₄)₆Cl₂, chlorapatite, Ca₅(PO₄)₃Cl, and spodiosite, Ca₂(PO₄)Cl. The immobilization process developed in this study involves a solid state process in which waste and precursor powders are mixed and reacted in air at temperatures in the range 700-800 °C. The ceramic products are non-hygroscopic free-flowing powders that only require encapsulation in a relatively low melting temperature phosphate-based glass to produce a monolithic wasteform suitable for storage and ultimate disposal.     

Low-temperature immobilization of actinides and other components of high-level waste in magnesium potassium phosphate matrices

The magnesium potassium phosphate (MPP) matrices formed by the room-temperature reaction of MgO and KH₂PO₄ solution were used for immobilization of simulated liquid alkaline high-level waste (HLW) containing actinides, fission and corrosion products. Novel procedures of solidification of HLW simulants were developed to increase stability of the MPP matrices to leaching radionuclides (Pu, Np, Am, Cs, Sr, Tc, I and Se), matrix-forming (K, Mg and PO₄) and admixture components (NO₃, NO₂, Na and others) as well as hazardous elements (Pb, Cr, Zn and others) according to the ANS, PCT, TCLP standards. Density (∼1.7 g/cm³), compressive strength (>20 MPa), radiation resistance of the matrices and chemical yield of radiolytic hydrogen (0.004 molecule H₂/100 eV) were determined. The phase composition of the matrices and distribution of radionuclides were studied by XRD and autoradiography correspondingly.     

Studies on immobilization of thorium in barium borosilicate glass

The barium borosilicate glass (BBS) matrix has shown considerable solubility of ThO₂ at 1000 °C. As seen by X-ray diffractometry (XRD) and Electron probe micro analysis (EPMA) up to 15.86 wt% of ThO₂ could be dissolved in this matrix. The homogeneity of thoria loaded glass was convincingly ascertained by EPMA. Attempts to load more than 16 wt% ThO₂ led to the phase separation of crystalline phases identified as major phase of ThO₂ and minor percentage of ThSiO₄ phase with altogether different morphologies, as seen by XRD. Interestingly, the back scattered images of thorite crystals point towards the presence of chemical zoning. The results being reported in this paper are of interest especially with respect to immobilization of other actinides in borosilicate glass matrix.     

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.     

X-ray absorption studies of bismuth valence and local environments in borosilicate waste glasses

X-ray absorption spectra (XAS) were collected and analyzed to characterize bismuth (Bi) environments in borosilicate glass formulations developed for the immobilization of high level nuclear wastes (HLW), from the bismuth phosphate process. Therefore, the structural role of Bi in these glasses is of interest; in addition in the present study, more particular interest in Bi originated from unusual foaming that was observed during melt cooling, where it was initially suspected that Bi³⁺ reduction to Bi⁰ may generate oxygen that caused the foaming. Observations from scanning electron microscopy of some HLW glass samples indicated a Bi-phosphate association. Bi L_III XAS of 13 Bi-containing waste glass formulations of various compositions were measured that exhibited varying degrees of melt foaming. The Bi XAS are similar for all glasses investigated, and indicate Bi³⁺O₃ nearest-neighbor environments with Bi–O distances near 2.13 Å. This environment is similar to the most localized Bi coordination characteristics in the crystalline Bi-silicates, eulytite (Bi₄Si₃O₁₂) and bismutoferrite (BiFe₂Si₂O₈OH). However, the Bi-environments in the glasses are distinctly different from the Bi-site in crystalline BiPO₄; therefore, XAS indicates no evidence of Bi-phosphate domains in the glasses measured. No XAS evidence was observed in any of the glasses investigated for Bi clustering, such as metallic Bi, or Bi–O–Bi bonding. Since the local Bi environments look similar for all glasses investigated, Bi XAS data and analyses show no association of the melt foaming problems with changes of Bi environments in the corresponding glass. The foaming may be dependent on chromate or phosphate behavior in the glass structure.     

Phasengleichgewichte in den systemen UO2-UO2,67-ThO2 und UO2 + x-NPO2

Solid-state chemical investigations have established that in the compositional range UO₂-UO_2.67-ThO₃ of the U-Th-O ternary system, the following single-phase domains exist: U₃O₈, which does not dissolve any ThO₂ in the solid state; an ordered M₄O₉ phase on the section between U₄O₉ and U₂Th₂O₉, below ≈ 1150 °C; and a phase with fluorite structure which occupies a large part of the system and which at 1250 °C is bounded by the compositions UO₂-UO_2.25 (U_0.43, ThO_0.57)O_0.25-ThO₃. The maximum O/M ratio of the “fluorite” phase is O:(U + Th) = 2.25. The highest oxidation valency of uranium is 5.30; this value falls as more thorium oxide is incorporated in the (U.Th)O_{2 + x} “fluorite” phase.     

模拟高放玻璃固化体的析晶行为

在高放废物玻璃熔制过程中,高放废物罐中心区域玻璃长时间处于高温状态,发生析晶现象。析晶增加了高放玻璃固化体的相界面,改变了其物理性质和化学性质。为了避免析晶的发生,必须研究高放玻璃固化体的析晶行为。本工作采用恒温热处理方式,研究模拟高放玻璃固化体的析晶行为。将模拟高放玻璃粉末分别在500、600、700、800、900℃下进行了0.5~96h的热处理。采用X射线衍射法(XRD)检测热处理后样品的结晶度和晶相。XRD测试结果表明:模拟高放玻璃固化体的最大结晶度析晶温度在700℃左右;热处理后产生了三类晶相辉石、斜长石和方铈矿,辉石出现在600~900℃,为主要晶相,方铈矿出现在700~900℃,斜长石只出现在700℃。     

Liquid Metal Extraction for Removal of Molybdenum from Molten Glass Containing Simulated Nuclear Waste Elements

Laboratory-scale experiments for removing Mo and MoO3 from molten borosilicate glass were performed using liquid Cu as an extractant. Removal of Mo from the simulated HLW glass containing oxides of Nd, Fe, Zr, Mo, Sn, Ni, Sr, Cd, Ru, and Se was also performed, and the fractions of these elements transferred into Cu were examined. Mixtures of Cu anda ternary SiO₂-B₂O₃-Na₂O glass containing metallic Mo or MoO₃ were heated in an alumina crucible at 1,673K in an Ar environment. The amounts of Mo and MoO₃ added to 10 g of the ternary glass were fixed at 0.1 and 0.15 g, respectively. As for the glass containing metallic Mo, more than 90% of Mo was extracted into liquid Cu. Spherical Cu metal buttons containing Mo formed on the bottom of the crucible when Cu was added at more than 10 times that of Mo on a mass basis. Removal of Mo from the glass containing MoO3 was also achieved by the addition of Si as a reducing agent for the reduction from MoO₃ to Mo. The fraction of Mo extracted into liquid Cu depended on the molar ratio of Si to Cu added to the glass. The fraction increased up to 84% with an increase in the molar ratio of Si/Cu. However, the excess addition of Si may enhance the chemical interaction between the metal phase and the glass phase, and some of the metal phase containing Mo remained in the glass phase without forming a metal button. The optimum molar ratio of Si/Cu that produces the highest removal fraction was found to be approximately 0.5. Almost the same removal fraction of 88% was obtained from the simulated HLW glass under the condition of Si/Cu = 0.5. Nearly 100% of Ru was extracted into Cu with Mo, while Sr, Zr, and Nd were hardly extracted and remained in the glass.