Ceramic matrices for plutonium disposition

One of the major issues related to the expanded use of nuclear power and the development of advanced nuclear fuel cycles is the fate of plutonium and “minor” actinides. In addition, substantial quantities of plutonium and highly enriched uranium from dismantled nuclear weapons now require disposition. There are two basic strategies for the disposition of the actinides: (1) to “burn” or transmute the actinides using nuclear reactors or accelerators; (2) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. This paper deals with actinide-bearing materials that support the latter approach. During the past two decades, a considerable amount of research and development has been done in an effort to develop matrices for the immobilization of plutonium and the “minor actinides”, Np, Am and Cm. A variety of waste form materials – oxides, silicates and phosphates – have been developed that have a high capacity for the incorporation of actinides, are chemically durable and, in some cases, resistant to the radiation-induced transformation to the aperiodic state. These waste forms can be selected depending on the composition of the waste stream that contains the actinides, the desired materials' properties of the waste form, and the geochemical and hydrologic conditions of the specific repository. The present state-of-knowledge for these materials is such that now one can design materials for very specific conditions, such as the thermal history and accumulated radiation dose, in a repository.     

Host phases for actinides in simulated metallic waste forms

Argonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists primarily of stainless steel cladding hulls containing undissolved metal fission products and a low concentration of actinide elements. This waste will be immobilized in a metallic waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). This paper presents transmission electron microscope, energy-dispersive X-ray spectroscopy, and electron diffraction observations of SS-15Zr alloys containing 2-11 wt% U, Np, or Pu. The major U- and Pu-bearing materials are Cr-Fe-Ni-Zr intermetallics with structures similar to that of the C15 polymorph of Fe₂Zr, significant variability in chemical compositions, and 0-20 at.% actinides. A U-bearing material similar to the C36 polymorph of Fe₂Zr had more restricted chemical variability and 0-5 at.% U. Uranium concentrations between 0 and 5 at.% were observed in materials with the Fe₂₃Zr₆ structure.     

地质环境中锕系元素水溶液化学研究进展

着重评述了近年来与高放废物深部地质处置有关的地质环境中锕系元素水溶液化学的研究进展，重点讨论了锕系元素在地下水中的溶解反应、水解反应、络合反应，还原反应，胶体的形成以及各种影响因素；并提出了近期我国应开展的若干研究课题     

Parametric Survey for Benefit of Partitioning and Transmutation Technology in Terms of High-level Radioactive Waste Disposal

Benefit of implementing Partitioning and Transmutation (P&T) technology was parametrically surveyed in terms of high-level radioactive waste (HLW) disposal by discussing possible reduction of the geological repository area. First, the amount and characteristics of HLWs caused from UO₂ and MOX spent fuels of light-water reactors (LWR) were evaluated for various reprocessing schemes and cooling periods. The emplacement area in the repository site required for the disposal of these HLWs was then estimated with considering the temperature constrain in the repository. The results showed that, by recycling minor actinides (MA), the emplacement area could be reduced by 17–29% in the case of UO₂-LWR and by 63–85% in the case of MOX-LWR in comparison with the conventional PUREX reprocessing. This significant impact in MOX fuel was caused by the recycle of 241Am which was a long-term heat source. Further 70–80% reduction of the emplacement area in comparison with the MA-recovery case could be expected by partitioning the fission products (FP) into several groups for both fuel types. To achieve this benefit of P&T, however, it is necessary to confirm the engineering feasibility of these unconventional disposal concepts.     

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO₂ and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO₃ content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO₂ and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of ²⁴¹Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.     

Glass-ceramic nuclear waste forms obtained by crystallization of SiO2-Al2O3-CaO-ZrO2-TiO2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th): Study of the crystallization from the surface

Glass-ceramic materials containing zirconolite (nominally CaZrTi₂O₇) crystals in their bulk can be envisaged as potential waste forms for minor actinides (Np, Am, Cm) and Pu immobilization. In this study such matrices are synthesized by crystallization of SiO₂-Al₂O₃-CaO-ZrO₂-TiO₂ glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th) as surrogates. A thin partially crystallized layer containing titanite and anorthite (nominally CaTiSiO₅ and CaAl₂Si₂O₈, respectively) growing from glass surface is also observed. The effect of the nature and concentration of surrogates on the structure, the microstructure and the composition of the crystals formed in the surface layer is presented in this paper. Titanite is the only crystalline phase able to significantly incorporate trivalent lanthanides whereas ThO₂ precipitates in the layer. The crystal growth thermal treatment duration (2-300 h) at high temperature (1050-1200 °C) is shown to strongly affect glass-ceramics microstructure. For the system studied in this paper, it appears that zirconolite is not thermodynamically stable in comparison with titanite growing form glass surface. Nevertheless, for kinetic reasons, such transformation (i.e. zirconolite disappearance to the benefit of titanite) is not expected to occur during interim storage and disposal of the glass-ceramic waste forms because their temperature will never exceed a few hundred degrees.     

Recovery of minor actinides from spent molten salt waste and decontamination of molten salt waste

Recovery of minor actinides from spent molten salt is one of the important issues. Decontamination of spent molten salt waste is also the problem to be solved for establishment of pyrochemical reprocessing. The decontamination method of spent molten salt waste with recovery of minor actinides has been proposed. Our proposed process is based on the hydrometallurgical process. This process consists of the following processes. First, the spent molten salt waste is dissolved in aqueous solution. Next, the minor actinides are recovered by chromatographic techniques using the pyridine resin in the methanolic hydrochloric acid solution. In the last process, the spent molten salt waste is decontaminated by the cation-exchange chromatography. In the present paper, the adsorption behavior of minor actinides, rare earth elements, alkaline earth elements, and alkali metal elements on pyridine resin is reported. The demonstration experiment of the recovery of the minor actinides from simulant spent molten salt waste is also reported.     

镅锔分离研究进展

乏燃料后处理产生的高放废液中Am和Cm是长期释热的主要来源,将它们分离出来并进一步进行分离和处置,对高放废物的长期安全处理处置具有重要意义。另外,超钚元素生产涉及Am和Cm材料的获取以及辐照后靶件中Am和Cm的化学分离。因此Am、Cm的分离一直是锕系元素化学与材料研究的重要领域之一。但是Am、Cm之间的分离相当困难,水溶液中Am、Cm基本均以正三价离子形式存在,化学性质非常相似。早期的离子交换法分离因子低,近年来主要研究将Am(Ⅲ)氧化到高价态实现分离,或通过Am、Cm与配体的亲和力差异、不同配体组合产生“推拉效应”以提高分离因子。本文综述了相关研究现状,概述了主要流程研发情况,并展望了该领域的研究趋势。     

乏燃料处置的必要性及其处置库环境化学行为

为了确保核燃料循环的安全性,不宜处理的乏燃料也应该同玻璃固化体一样作为高放废物进行深地质处置。本文综述了一些前期工作,归纳了空气侵入和水的辐解产生氧化性产物是导致乏燃料UO2基体氧化溶解的主要因素; 核燃料浸出实验结果显示铀和锕系镧系元素每天的浸出量是相应核素总量的1/107,比裂变产物的浸出速率小一个数量级。铁金属被各国选为高放废物处置容器材料的原因是其低价格、高强度和优秀的还原能力。在最不利的地下水侵入深地质处置库、近场处置容器防腐层破损的情景下,铁容器材料表面与地下水反应产生氢气,氢气通过还原反应消耗辐解产生的氧化性自由基和分子, 并能还原乏燃料表面的U（Ⅳ）,大幅度减缓乏燃料的腐蚀和溶解;乏燃料中裂变产物贵金属合金颗粒对氢气有催化作用;处置容器表面铁金属能还原沉积溶解的多价态核素U（Ⅵ）、Np（Ⅴ）、Tc（Ⅶ）、Se（Ⅳ）和Se（Ⅵ）。希望本文对我国确立以铁基金属为处置容器材料的包括乏燃料在内的高放废物深地质处置概念有参考作用。     

Toward On-Site Closed Nuclear Cycles Not Requiring Deep Burial of Waste

We discuss a non-chemical means for onsite reprocessing of spent fuel from hybrid reactors such as LIFE and also deep burn fission reactors. Using a plasma-based Archimedes Filter of standard design, actinides could be removed in a few passes through the Filter to qualify as TRU waste that could be disposed of in a site like WIPP. An improved Filter is discussed that could reduce waste to 1 cubic meter per year, suitable for shallow burial.     

高放废液中锕系离子分离研究进展 Ⅰ.双酰胺荚醚与锕系离子的配位化学

如何处理处置核电站反应堆产生的乏燃料及乏燃料后处理过程产生的高放废液是发展安全核能面临的一个主要问题。为提高核能的安全性、减少需要长时间深地层处置的高放废物量、有效利用地球上有限的可裂变材料资源,世界上发展核能的国家在过去几十年发展了从高放废液中分离少量锕系元素离子的萃取分离流程。近年来,双酰胺荚醚类化合物在锕系元素分离方面备受关注,本文从基础配位化学角度综述近期这类化合物与锕系元素离子相互作用等方面的研究结果。     

Recent achievements in the development of partitioning processes of minor actinides from nuclear wastes obtained in the frame of the NEWPART European programme (1996–1999)

Partitioning of long-lived minor actinides (americium and curium) from the nuclear wastes issuing the reprocessing of nuclear spent fuels, in order to transmute them into short-lived nuclides or to condition them into stable crystalline matrices, was the subject of intense research within the NEWPART research program of the European 4^th Frame Work Program, FWP (1996–1999). The target waste considered was the acidic raffinate (HAR) issuing the reprocessing of the used nuclear fuels by the PUREX process. A two step separation process based on liquid-liquid extraction was designed. The first step consists in the co-separation of the mixture of trivalent actinides and lanthanides from the HAR by extraction with a malonamide extractant (DIAMEX process), while the second step concerns the actinides(III)/lanthanides(III) group separation (SANEX process). Several DIAMEX and SANEX processes were developed and successfully tested with cold, spiked and genuine high active effluents. The research carried out also included basic and fundamental works in order to better understand the relationships between the structures of the extractants and their affinities for the target metal ions. The lecture highlighted both the basic and applied aspects of the research. This work will be pursued (PARTNEW program) within the 5^th FWP of the European Union during the period 2000–2003.     

Glass-ceramic nuclear waste forms obtained from SiO2-Al2O3-CaO-ZrO2-TiO2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th): study of internal crystallization

Glass-ceramic waste forms such as zirconolite (nominally CaZrTi₂O₇) based ones can be envisaged as good candidates for minor actinides or Pu immobilization. Such materials, in which the actinides (or lanthanides used as actinide surrogates) would be preferentially incorporated into zirconolite crystals homogeneously dispersed in a durable glassy matrix, can be prepared by controlled crystallization (nucleation + crystal growth) of parent glasses belonging to the SiO₂-Al₂O₃-CaO-ZrO₂-TiO₂ system. In this work we present the effects of the nature of the minor actinide surrogate (Ce, Nd, Eu, Gd, Yb, Th) on the structure, the microstructure and the composition of the zirconolite crystals formed in the bulk of the glass-ceramics. The amount of lanthanides and thorium incorporated into zirconolite crystals is discussed in relation with the capacity of the glass to accommodate these elements and of the crystals to incorporate them in the calcium and zirconium sites of their structure.     

新型锕系元素硼酸盐结构研究综述

近几年来,一系列新的锕系元素(钍、铀、镎、钚、镅、锔)硼酸盐化合物由硼酸熔融反应制备得到。这些化合物具有异常复杂的晶体结构以及非常优越的物理性质。硼酸钍具有纯无机多孔阳离子框架结构而拥有显著的阴离子交换性能,并能够选择性地将放射性核废液中的放射性核素99 Tc几乎完全提取出来,且保持高度的稳定性;硼酸铀化合物具有非常复杂的拓扑结构,且大部分结晶于非中心对称空间群;硼酸镎化合物经常显示出镎的混合价态,其中包含一个三重价态共存于同个化合物中的罕见例子,从而提供了一个崭新的轻锕系元素废料储放形式;硼酸钚提供了新奇的三价锕系元素配位环境;硼酸镅与硼酸锔同时显示出与硼酸钚与硼酸镧的显著差异,从而衍生出新的镧锕分离及锕系内部分离方略。迄今为止,关于超钚元素化合物的晶体结构与化学键的研究屈指可数,该系统研究具有非常显著的意义并且具有十足的挑战性。对此类化合物的进一步认识极有可能促进新一代核废料储放形式的探究及核燃料循环工艺的发展,进而阻止锕系放射性废物在环境中的扩散。     

Actinide recycling by pyro-process with metal fuel FBR for future nuclear fuel cycle system

Pyro-metallurgical technology is one of potential devices for future nuclear fuel cycle. Not only economic advantage but also environmental safety and strong resistance for proliferation are required for the fuel cycle. In order to satisfy the requirement, actinides recycling applicable to LWR and FBR cycles by pyro-process has been developed since more than ten years in CRIEPI. The main technology is electrorefining for U and Pu separation and reductive-extraction for TRU separation, which can be applied on oxide fuels through reduction process as well as metal fuels. The application of this technology on separation of TRU in HLLW through chlorination could contribute to the improvement of public acceptance on the geologic disposal.

The main achievements are summarized as follows:

• -|The elemental technologies, such as electrorefining, reductive extraction, injection casting and salt waste treatment and solidification, have been developed successfully with lots of experiments

• -|The fuel dissolution into molten salt and uranium recovery on solid cathode for electrorefining have been demonstrated by engineering scale facility in Argonne National Laboratory by using spent fuels and in CRIEPI by uranium tests.

• -|Single element tests, using actinides, showed the Li reduction to be technically feasible, remaining the subjects of technical feasibility on multi-elements system and on effective recycle of Li by electrolysis of Li₂O.

• -|Concerning on the treatment of HLLW for actinide separation, the conversion to chlorides through oxides has been also established through uranium tests.

• -|It is confirmed that more than 99% of TRU nuclides can be recovered from the high level liquid waste by TRU tests

• -|Through these studies, the process flow sheets for reprocessing of metal and oxide fuels and for partitioning of TRU separation have been established.

The subjects to be emphasized for further development are classified into three categories, that is, process development (demonstration), technology for engineering development, and supplemental technology.

The metal fuel FBR has a high potential for recycling actinides by integration with pyro-reprocessing. Alloys of U-Pu-Zr with minor actinides are investigated from points of fuel properties. The miscibility and other characteristics suggest that the maximum content up to ca. 5 wt% of minor actinides is allowable in the matrix. Nine pins of metal fuel including minor actinides are ready for irradiation at Phenix fast reactor.     

Recovery of actinides from actinide-aluminium alloys by chlorination: Part I

Pyrochemical processes in molten LiCl-KCl are being developed in ITU for recovery of actinides from spent nuclear fuel. The fuel is anodically dissolved to the molten salt electrolyte and actinides are electrochemically reduced on solid aluminium cathodes forming solid actinide-aluminium alloys. A chlorination route is being investigated for recovery of actinides from the alloys. This route consists in three steps: Vacuum distillation for removal of the salt adhered on the electrode, chlorination of the actinide-aluminium alloys by chlorine gas and sublimation of the formed AlCl₃. A thermochemical study showed thermodynamic feasibility of all three steps. On the basis of the conditions identified by the calculations, experiments using pure UAl₃ alloy were carried out to evaluate and optimise the chlorination step. The work was focused on determination of the optimal temperature and Cl₂/UAl₃ molar ratio, providing complete chlorination of the alloy without formation of volatile UCl₅ and UCl₆. The results showed high efficient chlorination at a temperature of 150 °C.     

“Used fuel” vectors and waste minimization strategies for VHTRS operating without refueling

Generation IV Very High Temperature Reactors (VHTRs) are well-known for their flexibility with respect to feasible fuel cycle options. In this paper, the LEU- and TRU-fueled VHTR configurations are analyzed accounting for their capabilities to attain an extended single-batch OTTO (Once-Through-Then-Out) mode of operation without intermediate refueling. The requirement of waste minimization is imposed as one of the design constraints defining possible system configurations and deployment strategies. The resulting “used fuel” vectors are examined considering anticipated disposal options as well as viability of fuel reprocessing. A Monte Carlo-deterministic analysis methodology has been implemented for coupled design studies of VHTRs with TRUs using the ORNL SCALE 5.1 code system. The developed modeling approach provides an exact-geometry 3D representation of the VHTR core details properly capturing VHTR physics. The presented analysis is focused on prismatic block core concepts for VHTRs. It is being performed within the scope of the U.S. DOE NERI project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTR configurations.     

Compactibility and sintering behavior of nanocrystalline (Th1−xCex)O2 powders synthesized by co-precipitation process

Thoria (ThO₂) based ceramic material is a versatile and very important matrix for immobilization of plutonium and other tetravalent actinides either as a burning or a deposition material for final disposal. The aim of this study was to investigate the influence of the actinide concentration (simulated with cerium), the fabrication conditions and the properties of the produced powders on the compactibility and sinterability of the final products. The (Th_1−xCe_x)O₂ powders with ceria concentration varying from 5 to 50 mol% were synthesized by co-precipitation method. The pellets were then compacted from calcined and ground powders at pressures varying from 250 to 750 MPa. The produced pellets had a homogenous grain size and sintered densities of 0.88% to 0.95% TD, respectively.     

Development of a treatment process and immobilization method for the volume reduction of uranium-bearing spent catalysts for final disposal

The uranium catalyst had been used in several industrial fields. The spent uranium catalyst became problematic radioactive waste awaiting a management strategy for the final disposal. This work studies a process to greatly reduce the volume of a spent uranium catalyst waste and the generation of a suitable waste form for final disposal. The process consists of several steps such as selective dissolution of the SiO₂ catalyst support, precipitation of dissolved silicon followed by its purification for release, treatment of uranium-laden wastewater generated during the process, and immobilization of the final uranium-bearing astes for disposal. Based on bench scale-level experiments, the process was confirmed to be effective to reduce the volume of the uranium catalyst waste. The final volume reduction yield obtained in this work was over 80% from the volume of the initial uranium catalyst waste. The radioactivity of the secondary wastes, namely, the recovered silica and effluent generated from the process, was confirmed to be sufficiently managed for clearance with meeting the discharge criteria in Korea. The process could achieve the maximum volume reduction of the uranium catalyst waste to be transferred to a disposal site, with the by-products from the process being released, meeting discharge criteria in view of both nuclear and non-nuclear environmental regulations.