Immobilization of Radioactive Wastes in Hydrothermal Synthetic Rock, (III)

The high-level radioactive waste form (21.8w/o waste, 47.7w/o α-quartz, 20.5w/o amorphous aluminosilicate, 10w/o Al(OH)₃) was produced with the addition of 10-N NaOH solution under hydrothermal hot-pressing conditions at 350°C and 66 MPa for 6 h. In order to characterize the waste form, following tests and measurements were performed: Soxhlet flow leach tests (97°C), static leach tests in deionized water at low temperatures (40, 90°C) and under hydrothermal conditions (100–350°C), crystalline phase determination, microstructure observation, compressive strength measurement, weight loss measurement by heat treatment and thermal conductivity measurement.

The waste form was mainly composed of α-quartz and had porous structure. As for leach rates determined by static low temperature leach tests, the leach rates of the waste form were much lower than a concrete form but slightly higher than a glass form. The waste form was stable under hydrothermal conditions in comparison with a glass form. It had high compressive strength and thermal resistance. Its thermal conductivity was higher at room temperature than that of a glass form.     

Surface Alteration of Pollucite under Hydrothermal Conditions

In connection with studies for evaluating the leachability of Cs from solidified radio-active waste, the surface alteration of pollucite was studied under hydrothermal conditions. Pollucite exposed to NaCl or KCl at 300°C formed on its surface a precipitate phase of analcime in the case of NaCl or leucite in the case of KCl. Cubo-octahedral analcime built up not only on the outermost surface of sample but also in the altered layer underneath. This crystal penetration was attributed to precipitation and isomorphous substitution of Na by Cs in the pollucite. The leachability of Cs proved to increase in keeping with salt concentration.

Exposure to CaCl₂ or to MgCl₂ formed a thick precipitate layer of anorthite in the case of CaCl₂ or clinochlore in the case of MgCl₂ over a thin reaction layer. This precipitate layer had the effect of inhibiting the migration of Cs. Si and Al to restrain Cs leaching into solution, which caused the amount of Cs leached to level off beyond 0.01 mol·dm^?3 salt concentration in the case of CaCl₂, and beyond 0.01 mol·dm^?3 in the case of MgCl₂.     

Surface Alteration of Sintered Product of Cesium Aluminum Silicate under Hydrothermal Conditions

Abstract

A sintered product of cesium aluminum silicate, CsAlSi₅O₁₂, formed an outer precipitated layer and a reaction layer when exposed to 3.0 M (= mol/dm³) NaCl under hydrothermal conditions above 250°C to 300°C. The precipitated layer consisted of crystallized phases of analcime and montmorillonite, and the reaction layer exhibited a concentration gradient for Cs. These layers grew with temperature and reaction time. When the solid sample was contacted with NaCl solutions below 0.1 M, the crystallized phase of pollucite was formed in the precipitated layer. Successive accumulation of pollucite layer contributed to the depression of leaching of Cs.     

Tungstate-based glass-ceramics for the immobilization of radio cesium

The preparation of tungstate-containing glass-ceramic composites (GCC) for the potential immobilization of radio cesium has been considered. The GCC materials were prepared by blending two oxide precursor compositions in various proportions. These included a preformed Cs-containing hexagonal tungsten bronze (HTB) phase (Cs_0.3Ti_0.2W_0.8O₃, P6₃/mcm) and a blend of silica and other oxides. The use of the HTB phase was motivated on the assumption that a HTB-based adsorbent could be used to remove cesium directly from aqueous high level liquid waste feeds. In the absence of the HTB, glass-ceramics were relatively easily prepared from the Cs-containing glass-forming oxide blend. On melting the mixture a relative complex GCC phase assemblage formed. The principal components of this phase assemblage were determined using X-ray powder diffraction, ¹³³Cs MAS-NMR, and cross-sectional SEM and included glass, various zeolites, scheelite (CaWO₄) and a range of other oxide phases and Cs-containing aluminosilicate. Importantly, under no circumstance was cesium partitioned into the glass phase irrespective of whether or not the composition included the preformed Cs-containing HTB compound. For compositions containing the HTB, cesium was partitioned into one of four major phases including zeolite; Cs-silica-tungstate bronze, pollucite (CsAlSi₂O₆), and an aluminosilicate with an Al/Si ratio close to one. The leach resistance of all materials was evaluated and related to the cesium distribution within the GCC phase assemblages. In general, the GCCs prepared from the HTB had superior durability compared with materials not containing tungsten. Indeed the compositions in many cases had leach resistances comparable to the best ceramics or glass materials.     

Candidate waste forms for immobilisation of waste chloride salt from pyroprocessing of spent nuclear fuel

Sodalite/glass bodies prepared by hot isostatic pressing (HIPing) at ～850 °C/100 MPa are candidates for immobilising fission product-bearing waste KCl–LiCl pyroprocessing salts. To study the capacity of sodalite to structurally incorporate such pyroprocessing salts, K, Li, Cs, Sr, Ba and La were individually targeted for substitution in a Na site in sodalite (Na vacancies targeted as charge compensators for alkaline and rare earths) and studied by X-ray diffraction and scanning electron microscopy after sintering in the range of 800–1000 °C. K and Li appeared to enter the sodalite, but Cs, Sr and Ba formed aluminosilicate phases and La formed an oxyapatite phase. However these non-sodalite phases have reasonable resistance to water leaching.Pure chlorapatite gives superior leach resistance to sodalite, and alkalis, alkaline and rare earth ions are generally known to enter chlorapatite, but attempts to incorporate simulated waste salt formulations into HIPed chlorapatite-based preparations or to substitute Cs alone into the structure of Ca-based chlorapatite were not successful on the basis of scanning electron microscopy. The materials exhibited severe water leachability, mainly in regard to Cs release. Attempts to substitute Cs into Ba- and Sr-based chlorapatites also did not look encouraging. Consequently the use of apatite alone to retain fission product-bearing waste pyroprocessing salts from electrolytic nuclear fuel reprocessing is problematical, but chlorapatite glass–ceramics may be feasible, albeit with reduced waste loadings. Spodiosite, Ca₂(PO₄)Cl, does not appear to be suitable for incorporation of Cl-bearing waste containing fission products.     

Distribution and Fixation of Cesium and Strontium in Zeolite A and Chabazite

The selective removal and fixation of Cs and Sr have been studied in zeolite A and chabazite. Cesium ion was preferentially distributed into chabazite with a high distribution coefficient (K _Cs>10³ cm³·g^?1) in the presence of NaCl (10^?1 mol·dm^?3). The K _Sr values for zeolite A attained about 10³ cm³·g^?1 in the pH range of 8～10, and they gradually decreased with an decrease in pH.

The initial rate of Cs adsorption was fairly fast in chabazite, and the adsorption ratio reached almost 100% within a few hours. The adsorption ratio of Sr in binderless A zeolite reached almost 100% after 15 h. The adsorption of Cs and Sr on these zeolites was followed by Langmuir-type isotherm. Cesium forms of these zeolites recrystallized to pollucite (CsAlSi₂O₆) above 900°C for zeolite A and above 1,200°C for chabazite. As for Sr forms, these zeolites changed to SrAl₂Si₂O₆ above 900°C.

These recrystallized phases were suitable hosts for the immobilization of Cs and Sr in the nuclear waste solutions.     

Temperature Effect on Plutonium Leach Rate of Nuclear Waste Glass

The two kinds of nuclear waste glass with similar composition, a ²³⁸Pu-doped and nonradioactive waste glass, were leached under the ISO-test conditions at temperature between 23 and 90°C. An activation energy of 22±10kJ/mole was obtained from the initial leach rates of Pu, which was much lower than the 78±9kJ/mole obtained from those of Si, Na, Sr and Cs, It is suggested that in the initial stages of leaching. Pu is not released from the waste glass with the same mechanisms as the releases of Si, Na, Sr and Cs, but the dissolution of hydrous plutonium dioxide PuO₂·_xH₂O formed on the glass surface becomes predominant. In the long duration tests (<32d), the release of Pu appears to be affected by the solubility of PuO₂·_xH₂O remaining in the leached surface layers.     

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.     

Reduction of Cesium Leaching Ratio from Cementitious Resin Forms Using Natural Active Silica

Cesium adsorption behavior of active silica, which is a natural acid clay composed of cristobalite and quartz, was evaluated for its applicability as Cs adsorbent to be added to cementitious waste forms containing spent ion exchange resin. Since active silica carried the Cs exchangeable silanol group (—SiOH) originally, the Cs distribution coefficient was remarkably high (<10⁴). It increased in saturated Ca(OH)₂ solution, simulating the cement paste, due to formation of new silanol groups. With its addition to the cementitious forms with ¹³⁴Cs adsorbed ion exchange resin solidified by slag cement, the Cs leaching ratio was reduced to below 1/10 that without active silica.     

Analysis of Leaching Behavior of Simulated LILW Glasses by Using the MCC-1 Test Method

Glasses developed for the treatment of low- and intermediate-level radioactive waste (LILW) from nuclear power plants were evaluated by using the Material Characterization Center-1 (MCC-1) leaching method. Tests were conducted at temperatures of 40, 70, and 90°C for three weeks in pH buffer solutions spanning the range from pH 4 to pH 11. Normalized mass losses and forward dissolution rates of major glass elements (B, Na, Al, Si, Co, Cs) were analyzed under each leaching condition. From these data, the forward rate equations depending on pH and temperature were defined using a nonlinear regression method. This equation provided an overall diagram of the leach rate with these parameters (i.e., pH and temperature). The forward dissolution rates of the glasses were found to have a V-shaped pH dependence. The glasses in the pH ranges were found to have a forward dissolution rate below 10 g/m²·d, when the temperatures were between 40 and 90°C and the leachant condition was pH 4–11. Except for the DG2 glass, the minimum forward dissolution rate (0.01–1 g/m²·d) was obtained at approximately pH 7–8. Compared with previously reported results, the developed glasses showed relatively high forward dissolution rates at the neutral region, while showing similar or lower rates compared with other glasses and ceramic waste forms at both extremes of pH.     

水泥固化体的铯的浸出行为

通过对硅酸盐水泥、碱矿渣水泥以及掺加沸石的碱矿渣水泥固化体在２５℃和７０℃下Ｃｓ￣＋的浸出行为研究发现，掺加３０％沸石的碱矿渣水泥固化体２８天Ｃｓ￣＋累积浸出量仅是硅酸盐水泥固化体的１／１０，是碱矿渣水泥固化体的１／６－１／４。几种水泥固化体Ｃｓ￣＋的累积浸出分数Ｐｔ与ｔ￣（１／２）作图都呈折线关系，前３天Ｃｓ￣＋扩散系数大于后期。Ｃｓ￣＋的浸出主要取决于水泥固化体的结构致密程度和固化体连通孔溶液中游离Ｃｓ￣＋浓度，而后者则又与水化产物束缚Ｃｓ￣＋的能力成反比和水化产物的溶解度成正比。因此碱矿渣水泥固化Ｃｓ￣＋的能力大于硅酸盐水泥。同时探讨了水泥固化体Ｃｓ￣＋的浸出模型。     

钾/钠离子对水热合成铯榴石结构的影响

水热合成铯榴石能低温、晶格固化放射性废物¹³⁷Cs,但在放射性废物中Cs⁺通常与同族的K⁺/Na⁺共存,两者可能会对水热合成铯榴石组成和结构产生影响。基于此,参考铯榴石化学计量组成CsAlSi₂O₆,首先利用偏高岭土、纳米二氧化硅和CsOH•H₂O为原料水热合成了铯榴石。然后分别以不同摩尔质量的KOH或NaOH取代CsOH•H₂O,研究了K⁺/Na⁺对水热合成产物组成和结构的影响。结果表明,铯榴石水热晶化温度相对较低,产物微观下呈多晶微球态;K⁺与Cs⁺共存时,体系水热产物主要为铯榴石微球,微球尺寸随K⁺掺入比例的提高而增加,但K⁺主要存在于晶化程度较低的钾铝硅酸盐水合物中,并未参与铯榴石形成;Na⁺与Cs⁺共存时,体系水热产物主要由铯榴石和方沸石(NaAlSi₂O₆•H₂O)组成,Cs⁺和Na⁺各自存在于铯榴石或方沸石结构中。铯榴石相对较低的水热晶化温度及其结构中Cs⁺不能迁移和进行离子交换的特性,使得水热条件下K⁺/Na⁺不能进入先形成的铯榴石结构中的碱金属离子位点。尽管如此,K⁺或Na⁺的存在可降低铯榴石的水热晶化温度、促进铯榴石晶粒生长,且当K⁺或Na⁺与Cs⁺共存时,Cs⁺仍优先进入并稳定固化于铯榴石晶体结构中。     

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.     

Accelerated Leach Tests of SRL-165 High-Level Waste Glass in Deionized Water

Accelerated short-term leach tests in a laboratory are neccessary in order to estimate, with reasonable accuracy, the long-term leaching behavior of high-level waste glass. In the present study, static leach tests of an SRL-165 high-level waste glass were carried out in deionized water at two different glass-surface-area to solution-volume ratios (SA/V-ratio), namely 0.85 and 0.079 cm^?1 at 90°C, and 0.85 cm^?1 at 40°C.

First, an equation was examined which related Si-concentrations with time, temperature and SA/V-ratio under the present static conditions. The parameter determined at 90°C, 0.85 cm^?1 can be used to calculate the Si-concentration at 40°C, 0.85 cm^?1. Second, at the low SA/V- ratio of 0.079 cm^?1, the concentrations of Ca and Mg in the leachates peaked and then decreased a little. The equation used above does not explain the variation of the concentrations of Ca and Mg at a low SA/V-ratio. The precipitation of Ca and Mg onto the glass surface is probably caused by the adsorption efficiency of the surface layer or the formation of crystalline materials at the low SA/V-ratio of 0.079 cm^?1. Third, the in-depth profiles of some elements obtained by secondary ion mass spectroscopy (SIMS) were qualitatively in agreement with the results of solution analyses. This indicates the particular usefulness of SIMS for analyzing leaching behavior of the glass in in-situ burial studies where solution analyses are often impractical.     

Solidification of “Cement-Glass”

Cement-glass, which is a mixture of sodium silicate (kNa₂O·mSiO₂nH₂O), silicon phosphate (P₂O₅·2SiO₂) and cement, was developed to solidify radioactive waste pellets in containers. The optimum molar ratio of Si(OH)₄, NaOH and P₂O₅2SiO₂ was found to be 2:2:1, based on compressive strength measurement of solidified samples. The compressive strength of solidified sodium silicate with silicon phosphate was increased when the amount of solidified SiO₂ per unit volume was increased by reducing the water content. Cement-glass had a higher Cs distribution factor than ordinary Portland cement.     

Alkali Hydrothermal Synthesis of Zeolites from Coal Fly Ash and Their Uptake Properties of Cesium Ion

Zeolites were synthesized from coal fly ash by hydrothermal treatment with KOH solutions. K-H zeolite (K₂Al₂Si₄O₁₂-nH₂O) was produced under optimum conditions of 160°C, 3 d, 1 M (=mol/dm³) KOH and liquid-solid ratio of 15cm³/g. The uptake behavior of radioactive cesium for the products was investigated by batch and column methods. The uptake equilibrium of Cs+ for the above product was attained within 2h yielding the distribution coefficient of above 10⁴ cm³/g. The uptake of Cs⁺ was followed by a Langmuir adsorption isotherm and the maximum uptake capacity was estimated to be 3.34 mmol/g. The successive removal of Cs⁺ was accomplished through the column packed with granular composites of product-alginate gel polymer.     

Study on the phosphate reaction characteristics of lanthanide chlorides in molten salt with operating conditions

A minimization of waste salt is one of the most important issues for the optimization of pyroprocessing. The separation of fission products in waste salts and the reuse of purified waste salt are promising strategies for minimizing the waste salt amounts. The phosphate precipitation of lanthanide is currently being considered for eutectic (LiCl–KCl) waste salt purification. In this research, the effects of molten salt temperature (400–550 °C) and reaction time (max. 180 min) upon conversion into the phosphate of lanthanides was investigated using 1 and 3 kg of eutectic salt. The conversion efficiency of lanthanides to molten salt-insoluble precipitates and phosphates was increased with an increase in molten salt temperature and operating time until it attained a specific temperature and time. K₃PO₄ as a precipitant was more favorable than Li₃PO₄ in terms of reactivity. To obtain over a 99% overall conversion efficiency, about 30 min was required in the case of using K₃PO₄ at 450 °C, but about 120 min in the case of using Li₃PO₄ at 550 °C. The lanthanide precipitates formed by a reaction with phosphate were a mixture of monoclinic structures, usually representing a polyhedron structure, and a tetragonal structure, representing a platelet structure.     

Accommodation of Cerium and Neodymium Oxides into Alumina

Alumina was studied on its applicability for a host material to incorporate lanthanide and actinide elements present in HLW. Alumina-based waste forms containing about 8 mol% in all of Ce₂0₃ and/or Nd₂0₃ were prepared by firing at 1,500°C for 25 h in a reducing atmosphere, after drying and denitrating the mixed solutions of aluminum, cerium and/or neodymium nitrates, and then pelletizing the powder mixtures under 36 MPa.

The obtained ceramic assemblages were analyzed by X-ray diffraction to determine crystalline phases produced. Tested also was the leachability of the assemblages in a nitric acid solution of pH=1 and deionized water at 150°C. These observations revealed that Ce prefers to be stabilized in magnetoplumbites, while Nd in perovskites. To obtain even denser assemblages, it seems necessary to suppress the formation of magnetoplumbites as much as possible. Leaching study showed that the relatively high dissolution rates of the forms in the solution of pH=1 g/10^?6g/cm²·d) were probably due to the innate reactivity of alumina with acid, and perovskites were more prone to leaching than magnetoplumbites. In the deionized water, the elemental release rates from the forms are lower than 10-⁷g/cm²·d, and not influenced practically by the presence of perovskites.     

Raman and X-ray absorption spectroscopic studies of hydrothermally altered alkali-borosilicate nuclear waste glass

Raman spectroscopy and X-ray absorption spectroscopy (XAS) are used to characterize structural changes that took place in hydrothermally altered (Na,K)-alumina-borosilicate glasses with different Na/K ratios, formulated as part of a durability study to investigate the behavior of glasses for nuclear waste storage. The hydrothermal experiments, or vapor hydration tests (VHT), were performed on each glass for 3 and 20 days at 200 °C to accelerate and approximate long-term alteration processes that may occur in a nuclear waste repository. Results found for both glasses and their VHT altered counterparts show little, if any, structural influence from the different starting Na/K ratios. X-ray diffraction, differential scanning calorimetry, scanning electron microscopy, and Raman spectroscopy indicate that the altered samples are mostly amorphous with small amounts of analcime-like and leucite-like crystals within 200 μm of the sample surface and contain up to 9.7 wt.% water or OH. The Raman data are nearly identical for the amorphous portions of all altered VHT samples investigated, and indicate that two glass structural changes took place during alteration: one, partial depolymerization of the alumina-borosilicate network, and two, introduction of water or OH. Al and Si XAS data indicate tetrahedral AlO₄ and SiO₄ environments in the original glasses as well as in the altered samples. Small energy shifts of the Si K-edge also show that the altered VHT samples have less polymerized networks than the original glass. Na XAS data indicate expanded Na environments in the VHT samples with longer Na-O distances and more nearest-neighbor oxygen atoms, compared with the original glasses, which may be due to hydrous species introduced into the expanding Na-sites.     

放射性废物固化体抗浸出性快速测定方法探讨

固化体的抗浸出性是放射性废物安全管理的一重要参数。目前，国内采用国标GB7023—86中的标准浸出试验方法测试固化体的抗浸出性，试验周期长。并且，国标GB14569.1—93仅对核素第42d的浸出率作了规定。这一规定不能很好反映不同固化基材、不同配方固化体间抗浸出性的差异。美国国家标准ANSI/ANS-16.1—2003采用快速浸出试验方法，并用浸出因子来表征核素的抗浸出性。本工作参照美国标准对试验结果的处理方法，对以往获得的真实或模拟放射性废物水泥固化体的浸出试验数据进行重新计算。计算结果表明，当核素累积浸出百分数小于20%时，核素的浸出率与浸出因子间存在一定的换算关系。据此，可考虑建立快速浸出试验方法和新的试验结果表述式，以较全面地判定放射性废物固化体的抗浸出性能。