Feasibility of Using Di-Dodecyl-Di-Octyl Diglycolamide for Partitioning of Minor Actinides from Fast Reactor High-Level Liquid Waste

The unsymmetrical diglycolamide, di-dodecyl-di-octyl diglycolamide (D³DODGA) is a modifier-free extractant proposed for partitioning of trivalent actinides from nitric acid medium. D³DODGA has been evaluated for the feasibility of using it in the absence of a phase modifier, for the partitioning of minor actinides from fast reactor high-level liquid waste (FR-HLLW). The extraction behavior of various metal ions present in the simulated FR-HLLW was studied in a solution of 0.1 M D³DODGA/n-dodecane from nitric acid medium. The distribution ratio of about 20 metal ions was measured as a function of concentration of nitric acid and other interfering ion. The extraction was found to be strongly dependent on the oxidation state of the metal ion. The extraction of Am(III) from 3–4 M nitric acid medium was quantitative in a single contact. However, it was accompanied by the quantitative extraction of fission products such as trivalent lanthanides (Ln(III)), Y(III), and Zr(IV). The extraction of Sr(II), Pd(II), and Ru(III) in 0.1 M D³DODGA/n-dodecane was not insignificant, but quite low. The extraction of Ba(II), Ni(II), Mo(VI), and Fe(III) was marginal and the extraction of Co(II), Sb(III), Mn(II), and Cs(I) in 0.1 M D³DODGA/n-dodecane was negligible. Our results indicated that 0.1 M D³DODGA/n-dodecane is a promising candidate for the separation of trivalent actinides from fast reactor high-level liquid waste containing significant quantities of trivalent lanthanides and actinides.     

Extraction of Radio-Strontium from Nitric Acid Medium Using Di-tert-Butyl Cyclohexano18-Crown-6 (DTBCH18C6) in Toluene–1-Octanol Diluent Mixture

An alternative extraction system to the SREX solvent using a diluent mixture comprising 4,4′(5′)di-tert-butylcyclohexano-18-crown-6 (DTBCH18C6) in 80% toluene–20% 1-octanol was developed and evaluated for Sr(II) extraction from pressurized heavy water reactor simulated high level waste (PHWR-SHLW). The acid uptake (5.7%) by the present solvent was significantly lower as compared to that by the SREX solvent (21%) which used 100% 1-octanol as the diluent. The extracted species conformed to the ion-pair [Sr(DTBCH18C6)(H₂O)_x]²⁺·2[(NO₃)(H₂O)_y]^?. Studies on Sr(II) extraction as a function of nitric acid concentration indicated more favorable extraction and stripping with the present solvent as compared to the SREX solvent. Loading studies with 0.025 M DTBCH18C6 in the diluent mixture, carried out using the Sr carrier, indicated a decrease in D_Sr from 3.1 with 10 ppm Sr carrier to 1.62 with 100 ppm Sr carrier. Other important physical parameters relevant for the extraction processes such as phase separation time (dispersion number), viscosity, and density were also measured. The radiation stability and reusability of the solvent was also investigated. In sharp contrast to the SREX solvent, with increasing absorbed dose the proposed solvent showed an increase in Sr extraction and an increased acid uptake.     

钙钛锆石固化处理高放射性废物的研究现状

钙钛锆石是主要的人造岩石固化基材之一,在固化处理从高放废物中分离得到的锕系元素方面受到广泛重视.本文对钙钛锆石本身的组成、结构和性能以及其合成及制备现状进行了较为全面的综述,并展望了其前景.     

模拟高放废液碱矿渣水泥固化体稳定性的研究

本文主要研究了碱矿渣水泥固化体的热力学稳定性、热稳定性和在地下水中的稳定性.研究结果表明,水化到28天时,固化体中处于介稳态的矿渣玻璃体基本消失,转化为稳定的水化产物,固化体的热力学稳定性良好;固化体在500℃的高温下仍具有足够的强度,热稳定性良好;同时,在地下水中,固化体也具有良好的稳定性.     

高稳定性SrF2-Na2O-Fe2O3-P2O5体系氟磷酸盐玻璃的制备与表征

⁹⁰SrF₂和高Na₂O、Fe₂O₃、P₂O₅含量的废物是两大类核电废物,将两者联合固化可获得以SrF₂、Na₂O、Fe₂O₃和P₂O₅为主要组成的玻璃固化体。本文采用X-射线荧光光谱仪、X-射线衍射仪、红外光谱仪、热膨胀仪、差热分析仪等研究了各样品的玻璃形成能力、实际组成、结构及热稳定性;采用溶解速率法和产品一致性测试方法研究了样品的耐水性。结果表明:1 000 ℃熔融保温0.5 h制得的玻璃中Sr损失较小而F质量损失达30%~34%;随着P₂O₅/SrF₂摩尔比的降低,玻璃的热稳定性急剧降低,而耐水性却有明显的改善;具有焦磷酸盐玻璃结构的样品稳定性好,其中SrF₂为30%(摩尔分数)的玻璃既具有高耐水性又具有较高的热稳定性。     

高放废液碱矿渣水泥固化体中矿渣反应程度研究

出于安全的考虑，高诳废液碱矿渣水泥固化体应该具有较高的热力学稳定性，要求在一个不太长的时间内使处于介稳态的矿渣玻璃体完全水化转化为热力学稳定的水化产物。为此分别用水杨酸甲醇－丙酮萃取法和非蒸发水法估测碱矿渣水泥中的矿渣反应率，用非蒸发水法结合ＤＴＡ法估测固化体中的矿渣反应率。     

Modeling Ion-Exchange Processing with Spherical Resins for Cesium Removal

The spherical Resorcinol-Formaldehyde and hypothetical spherical SuperLig® 644 ion-exchange resins are evaluated for cesium removal from radioactive waste solutions. Modeling results show that spherical SuperLig® 644 reduces column cycling by 50% for high-potassium solutions. Spherical Resorcinol Formaldehyde performs equally well for the lowest-potassium wastes. Less cycling reduces nitric acid usage during resin elution and sodium addition during resin regeneration, thereby significantly decreasing life-cycle operational costs. A model assessment of the mechanism behind “cesium bleed” is also conducted. When a resin bed is eluted, a relatively small amount of cesium remains within resin particles. Cesium can bleed into an otherwise decontaminated product in the next loading cycle. The bleed mechanism is shown to be fully isotherm-controlled vs. mass transfer controlled. Knowledge of residual post-elution cesium level and resin isotherm can be utilized to predict rate of cesium bleed in a mostly non-loaded column. Overall, this work demonstrates the versatility of the ion-exchange modeling to study the effects of resin characteristics on processing cycles, rates, and cold chemical consumption. This evaluation justifies further development of a spherical form of the SuperLig® 644 resin.     

铁磷熔体的氧化还原态和导电性能

用电导率电池测量了含钠铁磷模拟放射性核废料（HLW）熔体的电阻率,并分析了相应玻璃的Moessbauer谱,计算了铁离子的价态和氧化还原比,分析了温度、时间和氧化钠含量对熔体电阻率和玻璃氧化还原比的影响。结果发现：在Na2O含量低的熔体中,升温和降温过程的电阻率的变化是不可逆的：随着Na2O含量增加,不可逆性消失,熔体的电阻率随时间轻微下降。同时发现：Na2O3含量低的铁磷熔体的导电机理是电子性的,并且氧化还原比解释了其电阻率－温度曲线的不可逆性。铁磷模拟HLW熔体的电阻率类似于相近Na2O含量的钠铁磷系统熔体的电阻率。     

城市零星低放废液水泥固化技术

根据国家环保局下达的任务,为固化我国城市零星低放废液,选择了合适的水泥和外加剂。从固化体的抗压强度、电子显微镜、X射线衍射,差热分析和压汞测孔分析等实验结果说明,525矿渣水泥及外加剂C在干养护条件下,无论用自来水、PH>4的酸溶液或8％NaCO_3碱溶液调和,均能满足固化要求。当某些低放废液的酸度或碱度过高时,应将溶液预先中和后再固化。     

Use of Calix[4]-bis-2,3-naphthocrown-6 for Separation of Cesium from Pressurized Heavy Water Reactor Simulated High Level Waste Solutions (PHWR-SHLW)

Abstract

Distribution studies on Cs(I) were carried out from pressurized heavy water reactor (PHWR) simulated high level waste (SHLW) solution using calix[4]-bis-2,3-naphtho-crown-6 as the ligand. A mixture of 1:1 nitrobenzene and toluene was evaluated as a suitable diluent. The distribution ratio of Cs(I) increased with the aqueous feed acidity upto 3 M HNO₃ and decreased thereafter due to extraction of hydronium ions. The maximum D_Cs value at ～3 M HNO₃ suggested the possible application of the system for the recovery of radio-cesium from high level waste solution. The addition of 0.4% (v/v) Alamine 336 (a tertiary amine) facilitated the quantitative stripping of Cs(I) with distilled water. Quantitative extraction of Cs(I) from SHLW containing 0.32 g/L of Cs was observed in five contacts at O/A = 1/2 with 2.5 × 10^?3 M calix[4]-bis-2,3-naphtho-crown-6. Similarly, quantitative stripping of Cs(I) from the loaded organic phase was achieved in two contacts with distilled water at a volume ratio (O/A) of 2. Selectivity studies carried out using several radiotracers such as ¹⁴³Ce, ¹⁴⁰La, ¹⁴⁰Ba, ¹³⁷Cs, ¹⁰³Ru, ⁹⁹Mo, ^99mTc, ⁹⁷Zr, ⁹¹Sr, etc. indicated excellent selectivity for Cs. The reagent exhibited excellent chemical stability up to a period of six months.     

聚丙烯腈-亚铁氰化钾镍对Cs~+的动态吸附性能

通过动态吸附实验,研究了聚丙烯腈基亚铁氰化钾镍(PAN-KNi CF)对模拟放射性废水及实际放射性废水中Cs+的吸附效果,并采用SEM、FT-IR、XRD等对吸附剂进行了表征。结果表明,PAN-KNi CF能够有效地去除水中的Cs+,当溶液中含有共存离子时,PAN-KNi CF吸附Cs+的穿透曲线都从右向左移动;减容比为2 122时,PANKNi CF对实际放射性废水中137Cs的去污因子仍高达512;被PAN-KNi CF吸附的137Cs衰变成137Bam后会从PANKNi CF上解析下来。     

The application of in situ formed mixed iron oxides in the removal of strontium and actinides from nuclear tank waste

This article discusses the experimental program conducted at Argonne National Laboratory to study the performance of in situ formed mixed iron oxides (IS‐MIO) for the removal of strontium (Sr), plutonium (Pu), neptunium (Np), uranium (U), and americium (Am) from the Savannah River Site (SRS) radioactive tank waste. The boundaries for the experimental work were defined in collaboration with SRS. IS‐MIO was actually found to be a mixture of Fe(II) and Fe(III) oxides and hydroxides, including magnetite. Decontamination factor (DF) values were measured for both IS‐MIO and monosodium titanate (MST), the baseline sorbent used by SRS. DF values for IS‐MIO were found to be superior to MST for all isotopes studied. DF values for Pu, Np, and Sr, achieved within 30 min of IS‐MIO formation were orders of magnitude larger than the needed values. DF values for U and Am were less than the former three but still acceptable, and greater than MST. © 2010 American Institute of Chemical Engineers AIChE J, 2010     

玻璃固化国际现状及发展前景

介绍了玻璃固化国际应用状况,概述了玻璃固化工艺技术的发展,论述了4代玻璃固化熔制工艺,介绍了玻璃固化应用的新拓展,包括冷坩埚、就地玻璃固化、陶瓷玻璃固化、流动玻璃固化装置等新技术和新应用.评述了玻璃固化体安全处置的研究开发活动.     

Radionuclide Extraction by 2,6‐Pyridinedicarboxylamide Derivatives and Chlorinated Cobalt Dicarbollide

Abstract

The extraction properties of diamide derivatives of dipicolinamide (2,6‐pyridinedicarboxylamide or DPA, (R′R″NCO)₂C₅NH₃) in mixtures containing chlorinated cobalt dicarbollide in the acid form (HCCD), with and without a substituted polyethylene glycol (PEG), have been investigated. Distribution ratios of Cs, Sr, U, Eu and Am have been measured for various concentrations of diamide, PEG, aqueous phase nitric acid, with various HCCD:diamide ratios, and using different organic diluents. In the absence of HCCD, the diamides show little affinity for the extraction of Am or Eu from nitric acid solutions (distributions typically <1). Addition of HCCD with the diamide extractants indicates a pronounced synergistic effect with regard to actinide and lanthanide extraction; the observed Am and Eu distribution ratios typically increase by several orders of magnitude. Cesium is also appreciably extracted by the HCCD in the presence of the various diamides. Addition of PEG (to simultaneously extract Sr) with HCCD and diamide has minimal impacts on the Eu and Am distribution ratios. The initial data indicate that alkyl substituted DPA derivatives weakly affect the extraction properties with regard to actinides and lanthanides, while aryl substituents decrease extraction ability of the mixture. The results of this preliminary work indicate that numerous HCCD‐PEG‐DPA systems are promising and effective for the simultaneous extraction Cs, Sr, Am, and Eu from acidic solutions.     

超临界二氧化碳萃取放射性金属离子研究进展

介绍了超临界二氧化碳（SF-CO₂）在镧系、锕系等放射性金属离子萃取中的应用。综述了近年来该领域的最新研究成果。讨论了SF-CO₂应用于核废料以及混合废物的后期处理中的可能性。     

模拟高放射性废物碱矿渣水泥基固化体的性能研究

研究了用初压成型和振动成型工艺制造的碱矿渣水泥固化体的某些性能。结果表明，用碱矿渣水泥基材料固化模拟高放射性废物是可行的。采用初压成型工艺制造的碱矿渣水泥固化体，２８ｄ抗压强度达１１７．３ＭＰａ，孔隙率为９．６４％，且固化体还具有良好的抗冲击性能和较低的核素浸出率。     

沸石材料吸附水中放射性锶和铯的研究进展

放射性锶和铯是核工业生产过程中重要的核裂变产物,也是放射性废水中含量较多的放射性污染物,具有较长的半衰期和持续的生物毒性。本文综述了近几年国内外采用天然沸石、合成沸石和复合沸石吸附去除水中放射性锶、铯的研究进展。重点阐述了海藻酸盐、聚丙烯腈、炭材料以及磁改性金属与沸石构成的复合吸附材料对水中放射性锶、铯的吸附。复合沸石可以解决沸石粒径小、难分离、稳定性差等难题,强化了沸石的实用性。从吸附平衡时间、最大吸附容量、酸碱耐受度等方面归纳了沸石材料对水中放射性锶、铯的吸附特性,并分析总结了三类沸石的优缺点。最后,针对沸石材料在处理放射性锶、铯废水中的应用,展望了未来的研究方向,指出可以在降低原水浓度、开发组合工艺和加强模型模拟等方面的研究进行完善,从而推动沸石材料未来在工程技术领域的应用。     

模拟放射性废物泥浆陶瓷固化的烧结试验研究

采用钾长石、粘土、煅烧氧化铝为原料,对模拟放射性废物泥浆的陶瓷固化进行了研究。结果表明:该放射性废物泥浆陶瓷固化的最佳成型工艺为压制成型;烧结温度在1100℃以上的固化体抗渗性良好;本实验所选取的3个实验配方的最佳烧结温度范围均为1100-1150℃;K ,Na 在干燥过程中或烧结低温阶段的迁移扩散以及Na2SO4在 1200℃左右的分解,对固化体的致密烧结会产生不利的影响。