模拟锶铯废物铁磷酸盐玻璃固化及化学稳定性

研究了铁磷酸盐玻璃固化我国高放废液全分离流程中产出的锶废物及铯废物，熔制了相应的固化体。用XRD，IR测试了固化体的微观结构，用产品一致性试验方法(PCT)研究固化体的化学稳定性。研究表明：在所选的废物包容量范围内可熔制得均质模拟废物铁磷酸盐玻璃固化体，玻璃的主要结构基团为P2O7^4-，均质玻璃固化体有较好的化学稳定性。     

自蔓延高温合成固定放射性废物

以在氧化铬（CrO3）为氧化剂，采用自蔓延高温合成（SHS）技术制备了含锶离子（Sr^2 ）的钙钛矿（CaTiO3）化 的，并研究了反应产物的合成机理、微观组织和对Sr^2 的包容量。实验结果表明；以CrO3为氧化剂，在SHS反应过程中，可获得更高的反应温度，由于大量的液相存在，可使反应在瞬间完成，而且所获产物密度高、成分均匀的含锶CaTiO3；制备的CaTiO3可将核素固化于基体结构中，对SrO的包容量可达到35％（质量分数），从而达到封闭、隔离核素，防止放射性废物污染环境的目的。     

自蔓延高温合成法固化高放射性核废料

用新型氧化剂--三氧化铬(CrO3)自蔓延高温合成了含锶核素(Sr2 )的钙钛矿固化体,分析了自蔓延高温反应的理论,并用X射线衍射、扫描电子显微镜和粉末浸出等分析了合成产物的物相组成、微观结构及浸出率.结果表明:用CrO3作氧化剂,可提高反应的绝热燃烧温度,使合成反应完全在液相下完成,有利于生成致密均匀的固化体;固化体浸出率低,皆小于10mg/(m2?d),化学稳定性好,能长期稳定地处理高放射废物.     

“碱-矿渣-粉煤灰-煅烧高岭土”水合陶瓷固化模拟中放α废液

以钕和铈作为放射性铀和钚的模拟核素,采用碱激发和水热法制备了模拟中放α废液"碱-矿渣-粉煤灰-煅烧高岭土"水合陶瓷固化体。研究了固化体的结构组成、力学性能、抗侵蚀性和化学稳定性。结果表明:固化体的主要晶相组成为方沸石和NaPl沸石;在水热条件下,固化体中掺入模拟中放α废液后有利于沸石晶相的形成。抗压强度测试表明,固化体中废物包容量达到40%左右时仍能满足水泥固化体抗压强度的要求。固化体经硫酸盐介质和地下水侵蚀28 d后,其抗压强度损失较少,外观没有裂纹和龟裂,说明固化体具有较好的的抗硫酸盐和地下水侵蚀性能。另外,静态浸出测试结果表明,Nd和Ce的浸出率在10~(-8)~10~(-6)cm/d范围,说明固化体具有优良的化学稳定性。     

用钛硅酸盐玻璃固化模拟含铯废物的研究

设计了钛硅酸盐玻璃固化模拟含铯废物的配方,熔制了模拟含铯废物钛硅酸盐玻璃.用排水法测定了玻璃的密度,用XRD、IR等研究了玻璃的物相组成和结构,用PCT试验方法研究了玻璃固化体的化学稳定性.结果表明,当Na2O在Na2O-TiO2-SiO2三元体系中的量为24～34mol%,TiO2/SiO2小于0.58,配料中废物包容量小于22.74mol%时,体系的熔制温度较低,所得样品为密度较高、软化温度较高、化学稳定性较好的均质玻璃;TiO2/SiO2较大或废物包容量较高的配料,在形成玻璃的过程中容易失透,样品的化学稳定性差.     

基于核废物固化应用的铁磷酸盐材料研究进展

铁磷酸盐材料是一类新型的核废物固化材料,铁磷酸盐玻璃具有较高的化学稳定性和较低的熔融温度,独居石、磷灰石等是固化核素优选的晶体,铁磷酸盐玻璃陶瓷固化兼有玻璃基体对多种元素的包容性和陶瓷固化对特定放射性核素抗辐射能力的优势。综述了铁磷酸盐玻璃、陶瓷及玻璃陶瓷的特点及其在高放废物固化方面的研究现状及进展,比较了用于核废物固化的铁磷酸盐固化体的体系及特点。     

钛酸盐陶瓷固化Sr2+核素的SHS热力学

探索采用燃烧合成(SHS)技术制备包容锶核素(Sr2 )的钛酸盐陶瓷固化体,根据自蔓延高温合成热力学分析,计算钛酸盐固化体的绝热燃烧温度Tad大于临界绝热燃烧温度,说明反应能自发进行并合成致密的CaTiO3固化体。     

核屏蔽材料铅硼聚乙烯高温熔融处理研究

核辐射屏蔽材料铅硼聚乙烯对中子和γ射线均具有较好的屏蔽效果,因此在核反应堆辐射防护中使用广泛.使用过程中被核素污染的铅硼聚乙烯属于核废物,需要特殊处理.采用高温熔融玻璃固化的方式可以实现对铅硼聚乙烯的减容及固化处理.通过高温熔融的方式处理铅硼聚乙烯模拟核废物,研究结果表明,铅硼聚乙烯模拟核废物在900℃下即可熔融形成铅硼玻璃.然而产品一致性测试结果表明,该铅硼玻璃的核素及硼浸出值较高.通过引入一定量的玻璃添加剂(SiO2),其玻璃化学稳定性显著提高,达到核废物玻璃固化体化学稳定性要求.     

模拟高钠高放废物铁硼磷酸盐玻璃固化体的结构和性能

以36Fe2O3--10B2O3--54P2O5为基础玻璃,制备了不同模拟高钠高放废物包容量的铁硼磷酸盐玻璃固化体,用Fourier变换红外光谱测试方法系统研究了由废物包容量引起的玻璃固化体结构变化,并用溶解速率法初步测试了固化体的化学性能。结果表明:随着废物包容量的增加,固化体试样结构中(PO4)3-四面体基团增加,[BO3]基团向[BO4]基团转变,磷酸盐基团彼此间的连接程度减小,Fe—O—P键在包容量为25%(质量分数)到30%时存在量较大。玻璃固化体网络结构以(PO4)3-四面体基团为主,易水化的(PO3)-磷酸盐基团的含量很小。但固化体结构中[BO3]基团的存在量还较大,该组分的基础玻璃网络形成体氧化物配比还可进一步优化。当废物包容量小于40%时,固化体不同浸泡周期的质量损失速率均在10--8 g/(cm2·min)数量级。     

GdEuZrCeO_7烧绿石陶瓷核废物固化体的制备及化学稳定性

选择镧系元素Eu、Ce分别模拟高放废物中的三价和四价锕系核素,以硝酸盐和氧化物为原料,分别采用普通烧结法和液相烧结法制备了GdEuZrCeO_7烧绿石陶瓷核废物固化体。利用XRD、SEM-EDS对固化体的物相组成、微观结构和化学组成进行了测试表征,研究了烧结方法、烧结温度和烧结助剂对固化体物相组成、晶粒大小以及致密性的影响,并采用PCT静态浸出法研究了固化体的化学稳定性。结果表明:GdEuZrCeO_7烧绿石陶瓷固化体的物相结构为无序缺陷萤石型结构;较高的烧结温度有利于固化体结晶度和致密性的提高,液相烧结和添加烧结助剂对固化体致密性的提高并不明显;PCT结果说明固化体具有较好的化学稳定性,元素的归一化浸出率总体处于10~(-7)～10~(-4) g·m~(-2)·d~(-1)范围。     

Pyrochlore glass-ceramics for the immobilization of molybdenum-99 production wastes: Demonstrating scalability and flexibility to waste stream variance

Pyrochlore glass ceramics have been fabricated via in-situ crystallization under reducing conditions by both sintering and hot isostatic pressing (HIPing) as candidate wasteforms for the acidic waste biproduct of Mo-99 radiopharmaceutical production. The tailored wasteform demonstrates flexibility in the wasteform design to receive the required waste variability, it also suitably passes high-level waste performance requirement, and successfully scales to 1 kg scale with 45 wt.% waste loading. U-rich pyrochlore as the major phase was confirmed by X-ray diffraction, scanning electron microscopy and energy dispersive X-ray spectroscopy, with residual glass and minor secondary phases. The presence of both U⁴⁺ and U⁵⁺ valences in the wasteforms was revealed by diffuse reflectance spectroscopy. Addition of glass content had little influence on the pyrochlore composition but facilitated minor perovskite formation. The up-scaled dense, HIPed sample showed elemental releases of < 2 g/L for all elements in durability experiments.     

Hot Isostatically Pressed (HIPed) fluorite glass-ceramic wasteforms for fluoride molten salt wastes

Molten pyroprocessing salts can be used to dissolve used nuclear fuel from a reactor allowing recovery of the actinides. Previously, ANSTO have demonstrated hot isostatically pressed (HIPed) sodalite glass-ceramic wasteforms for eutectic (Li,K)Cl salts containing fission products, but this system cannot be used for the analogous molten alkali fluoride salts (eg, FLiNaK), which have utility in the application of the next generation of nuclear reactors. In this work, a novel glass-ceramic composite wasteform has been prepared by HIPing, as a candidate for the immobilization of fission product-bearing FLiNaK salts. The wasteform has been tailored to immobilize the high fluoride content of the waste within fluorite, whereas the waste alkali elements are incorporated in a durable sodium aluminoborosilicate glass, with total waste loadings of ~17-21 wt% achieved. It was also demonstrated that the speciation of Mo- and Sb-simulated fission products was altered by adding Ti metal due to a controlled redox environment. The resulting candidate wasteform has been studied by X-ray diffraction and scanning electron microscopy, including the HIP canister-wasteform interaction zone, and its performance assessed via leaching studies using the PCT and ASTM C1220 leaching protocols. Dr Vance very much enjoyed the challenge of wasteform design for problematic nuclear wastes, for which fission product-bearing FLiNaK salts are a clear example. His ability to hone in on a wasteform solution with viable waste loadings that meet performance requirements was testament to his nearly 40 years experience in nuclear waste immobilization. The samples discussed in this work represent the last wasteform materials that he prepared.     

Profiling hot isostatically pressed canister–wasteform interaction for Pu-bearing zirconolite-rich wasteforms

Zirconolite-rich full ceramic wasteforms designed to immobilize Pu-bearing wastes were produced via hot isostatic pressing (HIP) using stainless steel (SS) and nickel (Ni) HIP canisters. A detailed profiling of the elemental compositions of the major and minor phases over the canister–wasteform interaction zone was performed using scanning electron microscopy combined with energy-dispersive X-ray spectroscopy (SEM-EDS) characterization. Bulk sample analyses from regions near the center of the HIP canister were also conducted for both samples using X-ray diffraction and SEM-EDS. The sample with the Ni HIP canister showed almost no interaction zone with only minor diffusion of Ni from the inner wall of the canister into the near-surface region of the wasteform. The sample with the SS HIP canister showed ∼100–120 μm of interaction zone dominated by high-temperature Cr diffusion from canister materials to the wasteform with the Cr predominantly incorporated into the durable zirconolite phase. We also examined, for the first time, changes to the HIP canister wall thickness caused by HIPing and demonstrated that no canister wall thinning occurred. Instead, in the areas examined, the canister wall thickness was observed to increase (up to ∼20%) due to the compression occurring during the HIP cycle. Further, only sparse formation of (Cr, Mn)-rich oxide particles were noted within the HIP canister inner wall area immediately adjacent to the ceramic material, with no evidence for reverse diffusion of ceramic materials. Though the HIP canister–wasteform interaction extends to ∼120 μm when using an SS HIP canister for the system investigated, this translates to <<1 vol.% for an industrial scale HIPed wasteform. Importantly, the HIP canister–wasteform interactions did not produce any obviously less durable phases in the wasteform or had any detrimental impact on the HIP canister properties.     

SHS of mineral-like ceramics for immobilization of radioactive nuclear wastes

Synthetic analogs of titanate ceramics (perovskite and zirconolite) designed for use as a matrix for immobilization of high-level nuclear fuel reprocessing wastes (HLW). Such ceramics have been prepared by SHS method from a mixture of titanate ceramics and non-radioactive model oxides. Synthesis conditions have been optimized. The synthesized low-porosity cylindrical compacts exhibited a high strength and low rate for leaching Cs, Sr, Y, Ce, and La in bidistilled water. The phase composition and microstructure of synthesized products have been characterized. The immobilization of Cs was found to be accompanied by a marked loss of this element. The text was submitted by the authors in English.     

Immobilization of Sodium-Bearing High-Level Radioactive Waste in Synroc Containing (Na0.5Nd0.5)TiO3-Type Perovskite

This study on the immobilization of high-sodium-bearing HLW in synroc indicates that (Na_0.5Nd_0.5)TiO₃-type perovskite can be used to incorporate a high content of sodium in synroc. Synroc samples containing 13.0 wt% waste oxide and 5.7 wt% Na₂O show very well chemical durability and physical properties. The standard Synroc-C formulation can incorporate only 2 wt% Na₂O, so this study greatly improved the immobilization ability of sodium in Synroc-related material.     

Synroc technology: Perspectives and current status (Review)

Dr Eric (Lou) Vance spent 32 years at the Australian Nuclear Science and Technology Organisation (ANSTO), where he was dedicated to the development of Synroc technology, a waste treatment solution for intractable nuclear wastes. The original form of Synroc, a multiphase ceramic wasteform based on stable and leach resistant titanate minerals, was invented by Australian scientists in the late 1970s. This formulation was directed toward the immobilization of PUREX wastes from the reprocessing of nuclear fuels. Synroc at ANSTO under the scientific leadership of Dr Vance since evolved beyond these original titanate ceramics into a waste treatment technology platform. This platform can be applied to produce glass, glass-ceramic and ceramic wasteforms and offers distinct advantages in terms of waste loading and suppressing volatile losses. The platform therefore provides an opportunity to treat those waste streams that are problematic for glass matrices alone or existing vitrification process technologies. Such wastes include, for example, actinide-bearing wastes, those that contain large proportions of refractory elements, those with significant fission product or corrosive volatile emissions and those wastes resulting from radiopharmaceutical production. The implementation of the latter will see the industrialization of Synroc technology via a first-of-a-kind Synroc Waste Treatment Facility that is currently under construction at ANSTO. This paper will review Synroc technology, particularly noting the substantial and essential contributions from the late Dr Vance. The review will also provide some perspective on the development of the technology for nuclear waste immobilization and describe the significant recent advancements at ANSTO.     

CAUSTIC LEACHING OF HIGH-LEVEL RADIOACTIVE TANK SLUDGE: A CRITICAL LITERATURE REVIEW

ABSTRACT

The Department of Energy (DOE) must treat and safely dispose of its radioactive tank contents, which can be separated into high-level waste (HLW) and low-level waste (LLW) fractions. Since the unit costs of treatment and disposal are much higher for HLW than for LLW, technologies to reduce the amount of HLW are being developed. A key process currently being studied to reduce the volume of HLW sludges is called enhanced sludge washing (ESW). This process removes, by water washes, soluble constituents such as sodium salts, and the washed sludge is then leached with 2–3 M NaOH at 60–100°C to remove nonradioactive metals such as aluminum. The remaining solids are considered to be HLW while the solutions are LLW after radionuclides such as ¹³⁷Cs have been removed. Results of bench-scale tests have shown that the ESW will probably remove the required amounts of inert constituents. While both experimental and theoretical results have shown that leaching efficiency increases as the time and temperature of the leach are increased, increases in the caustic concentration above 2–3 M will only marginally improve the leach factors. However, these tests were not designed to validate the assumption that the caustic used in the ESW process will generate only a small increase (10 Mkg) in the amount of LLW; instead, the test conditions were selected to maximize leaching in a short period and used more water and caustic than is planned during full-scale operations. Even though calculations indicate that the estimate for the amount of LLW generated by the ESW process appears to be reasonable, a detailed study of the amount of LLW from the ESW process is still required. If the LLW analysis indicates that sodium management is critical, then a more comprehensive evaluation of the clean salt process or caustic recycle would be needed. Finally, experimental and theoretical studies have clearly demonstrated the need for the control of solids formation during and after leaching.

     

复相陶瓷(0.5BaO0.5SrO)[(1-y)TiO2yNb2O5]的介电性能研究

通过对组成进行一定的过量控制，利用传统陶瓷工艺，成功制备了具有钙钛矿相钛酸锶钡固溶体和钨青铜相铌酸锶钡固溶体两相稳定共存的(0.5BaO0.5SrO)[(1-y)TiO2yNb2O5](BSTN)复相陶瓷.用Riguaku D/max-2400型X射线衍射仪和KEITHLEY 3330阻抗分析仪对所得样品的相结构状态及介电性能进行了测试.结果表明：Y值在0.1～0.7之间的BSTN复相陶瓷为钙钛矿相和钨青铜相两相稳定共存的复合相；在该范围以外，则为单一相的钛酸锶钡与Nb或铌酸锶钡与Ti的固溶体.复相陶瓷的介电性能取决于钙钛矿相和钨青铜相的相对含量及其固溶度，当组成y约为0.7时，介电常数达到最大值；随测试频率的提高，介电常数和损耗降低.复相陶瓷中钨青铜相的铁电/顺电相转变温度与钨青铜相中固溶TiO2量相关，随固溶量增加，转变温度降低.同时，受复相陶瓷中钙钛矿相的影响，随钙钛矿相相对含量的增加，转变温度升高；钨青铜相的最低铁电/顺电相转变温度出现在约110℃，比Sr0.5Ba0.5Nb2O6下降约240℃.     

Current Understanding and Remaining Challenges in Modeling Long-Term Degradation of Borosilicate Nuclear Waste Glasses

Chemical durability is not a single material property that can be uniquely measured. Instead, it is the response to a host of coupled material and environmental processes whose rates are estimated by a combination of theory, experiment and modeling. High-level nuclear waste (HLW) glass is perhaps the most studied of any material yet there remain significant technical gaps regarding their chemical durability. The phenomena affecting the long-term performance of HLW glasses in their disposal environment include surface reactions, transport properties to and from the reacting glass surface, and ion exchange between the solid glass and the surrounding solution and alteration products. The rates of these processes are strongly influenced and are coupled through the solution chemistry, which is in turn influenced by the reacting glass and also by reaction with the near-field materials and precipitation of alteration products. Therefore, those processes must be understood sufficiently well to estimate or bound the performance of HLW glass in its disposal environment over geologic time scales. This article summarizes the current state of understanding of surface reactions, transport properties and ion exchange along with the near-field materials and alteration products influences on solution chemistry and glass reaction rates. Also summarized are the remaining technical gaps along with recommended approaches to fill those technical gaps.