Feasibility studies on the selective separation of fission palladium(II) by isoHex-BTP/SiO2-P adsorbent from HLLW

Aiming at the selective recovery of fission palladium(II) from high-level liquid waste (HLLW), the silica/polymer (SiO₂-P)-based isoHex-BTP adsorbent (isoHex-BTP/SiO₂-P) was synthesized by impregnating complexing agent isoHex-BTP into the multiporous SiO₂-P inert support. The feasibility of separation of Pd(II) from HLLW by isoHex-BTP/SiO₂-P was evaluated by batch experiment method. The results showed that isoHex-BTP/SiO₂-P exhibited much higher adsorption selectivity for Pd(II) than the other fission products, even Am(III) and Pu(IV) presented in HLLW. The ideal nitric acid concentration for the adsorption of Pd(II) by the adsorbent was shown to be ≥2 mol dm^–3. The adsorption of Pd(II) fits well to the pseudo-second-order kinetic model and Langmuir isotherm model. Quantitative Pd(II) desorption was achieved by using 0.5 mol dm^?3 SC(NH₂)₂ - 0.1 mol dm^?3 HNO₃ solution.     

Current status and future plans of Advanced ORIENT Cycle strategy

For minimization of the ecological risks inherent in nuclear fuel recycling, a new fuel cycle paradigm was proposed and its key technology developments have been carried out as a part of the Advanced Optimization by Recycling Instructive Elements (Adv.-ORIENT) Cycle strategy. The basic concept of the Adv.-ORIENT Cycle uses a three-pronged approach, separation, transmutation and utilization of nuclides and elements, based on the FBR fuel cycle. Fundamental research studies done in Adv.-OEIENT Cycle [Phase-1: 2006-2010] have led to the following findings.

1.

Cs and Sr separation and its utilization: Silica gel loaded with ammonium molybdophosphate (AMP) and hybrid organic microcapsules with crown ether D18C6 were investigated to chromatographically separate Cs and Sr, respectively. In particular, uptake experiments of Cs from solution simulating the spent fuel solution obtained were carried out by a batch method, and the uptake rates achieved were more than 90%.

2.

Minor actinide (MA)/lanthanide/fission product (FP) separation: A tertiary pyridine type resin (TPR) can be used to recover Am, Cm and lantanide elements with a high separation factor by a chromatographic method from spent fuel solution. The TPR can be used with hydrochloric acid (HCl) as well as nitric acid (HNO₃).

3.

Ru, Ph, Pd and Tc separation: A catalytic electrolytic extraction (CEE) method can effectively separate Ru, Ph, Pd and Tc. High recovery ratios of Ru, Rh, Pd, Tc, Se, etc. were achieved using HCl solutions. Rh co-deposition significantly accelerated reduction of Ru, Tc and Re using HNO₃ solutions.

4.

Ru, Ph, Pd and Tc utilization: Based on the mixed deposit obtained from the CEE experiments, Ru/Rh/Pd/Tc(Re)-Pt electrodes provided the highest catalytic reactivity in the electrolytic production of hydrogen in an alkali solution.

5.

Basic engineering research: Results of corrosion experiments showed that Hastelloy-B and Ta had a good anti-corrosive property in a wide range of HCl concentrations. The basic thermo-chemical stability of TPR and tri-butyl phosphorous (TBP, as a reference) was also experimentally studied, and the process safety conditions to be specified for practical use of TPR could be identified.

     

高放废液除锶技术的研究进展

高放废液的处理处置是影响核能可持续发展的重要因素。从高放废液中分离出⁹⁰Sr不仅能实现高放废液的安全处置,也具有较高的经济意义。本文在早期研究的基础上,结合近些年的研究情况,从沉淀法、离子交换法、溶剂萃取法、色层分离法和膜分离法5个方面介绍了从高放废液中分离⁹⁰Sr的方法,并进行了讨论。最后指出目前研究工作中存在的不足,并对其前景进行了展望。     

我国高放废液中铯分离研究进展

由于高放废液的放射性强、毒性大、组成复杂，从高放废液中分离铯是一个世界性难题。多年来国内外研究者一直在探索研究从高放废液中分离铯的方法，开发适合工业应用的铯分离技术，以解决从高放废液中分离铯的难题。一方面，我国现存的生产堆高放废液，浓缩倍数大、盐分高、放射性强，长期贮存风险大，需要进行妥善处理；另一方面，随着我国核电的快速发展和民用核燃料后处理的工业化，动力堆高放废液的处理问题也日益突出。针对这些需求，我国科技工作者们开展了大量从高放废液中分离铯的研究工作，取得了系列研究成果。近几十年来我国主要开展了离子交换、萃取色层和溶剂萃取分离高放废液中铯的研究，先后开发了亚铁氰化钛钾离子交换分离工艺以及杯芳烃冠醚萃取分离工艺，并进行了热实验验证以及台架实验。杯芳烃冠醚从高放废液中萃取分离铯的工作不但具备了工程应用的技术条件，也走在了世界前列。     

A study on selective precipitation of U(VI) by hydrophilic cyclic urea derivatives for development of a reprocessing system based on precipitation method

Selective precipitation ability of 2-imidazolidone (EU) and tetrahydro-2-pyrimidinone (PU) for U(VI) species in HNO₃ solutions containing U(VI), U(IV) (simulant of Pu(IV)), and simulated fission products (FPs) was investigated. As a result, it was found that these compounds precipitate almost quantitatively U(VI) as UO₂(NO₃)₂L₂ (L = EU, PU) from 3.0 M HNO₃ solution. In contrast, these urea derivatives form neither solid precipitates nor oily products with U(IV) in HNO₃ solutions containing only U(IV) species and even in U(VI)–U(IV) admixture system. Therefore, the separation of U(VI) from U(IV) was demonstrated to be achieved in use of EU and PU. Furthermore, EU and PU are capable to remove most of simulated FPs[Sr(II), Ru(III), Rh(III), Re(VII) La(III), Ce(III), Pr(III), Nd(III), and Sm(III)] from U(VI) to give their decontamination factors (DFs) higher than 100, while those values of Zr(IV), Mo(VI), Pd(II), and Ba(II) are necessary to be improved in both systems. From these results, it is expected that EU and PU are the promising precipitants for selective separation of U(VI) from HNO₃ solutions dissolving spent FBR fuels.     

萃取分离法处理高放废液的进展

评述了近几年用萃取分离法从高放废液中去除超铀锕系元素的进展情况，着重介绍世界上已有的应用前景较好的ＴＲＵＥＸ流程（美）、ＤＩＡＭＥＸ流程（法）、ＤＩＤＰＡ流程（日）、ＣＴＨ流程（瑞典）和ＴＲＰＯ流程（中国）。     

Continuous separation of molybdenum and zirconium from simulated high-level liquid waste with a Taylor–Couette contactor

Reprocessing of spent nuclear fuels generates high-level liquid waste (HLLW) which undergoes vitrification into borosilicate glass before final geological disposal. To ensure the quality of the glass, control of the concentration of chemical species such as molybdenum (Mo), which has an adverse impact on the vitrification process, is critical. Also, zirconium (Zr) can cause crud in washing process and Zr-93 is a long-lived fission product needed to be separated. In this study, a liquid–liquid countercurrent centrifugal contactor with Taylor–Couette flow (TC contactor) was applied to practical multi-species cases. Continuous separation of Mo and Zr from a simulated HLLW with bis(2-ethylhexyl) phosphoric acid (HDEHP) as extractant has been performed. Among a variety of metals in simulated HLLW, Mo, Zr, Y, and Fe are extractable, Mo and Zr were separated from HLLW by equilibrium, and Fe/Y separation was achieved by the effect of non-equilibrium state in TC contactor. Addition of tributyl phosphate could improve extraction of Mo. This study has expanded the scope of the TC contactor to multi-species separation processes.     

Selective Uptake and Recovery of Palladium by Biopolymer Microcapsules Enclosing Cyanex 302 Extractant

Selective uptake and recovery of platinum group metals are an important subject in the high level liquid waste treatment. An extractant having a strong affinity for palladium, bis(2,4,4-trimethylpentyl) monothiophosphinic acid (Cyanex 302, HA), was enclosed into microcapsules by utilizing highly immobilizing ability of biopolymer (alginate and alginic acid gel polymers). The uptake of Pd^2plus, Ru(NO)³⁺ and Rh³⁺ for granular microcapsules has been studied by batch and column methods. A relatively large distribution coefficient of Pd^2plus, K_d,Pd, of above 10⁴ cm³/g was obtained in the presence of 0.2–0.5 M HNO₃ and the separation factors of Pd/Ru and Pd/Rh were estimated to be above 10². The uptake of Pd²⁺ on microcapsules followed a Langmuir-type uptake isotherm and the maximum uptake capacity was estimated to be 0.72–0.96 mmol/g. These metal cations were chromatographically separated through the column packed with HA-alginate microcapsules.     

Optimization for removal of ruthenium from nitric acid solution by volatilizing with electrochemical oxidation

Ruthenium is a major fission product element among the platinum group elements (PGEs) in high-level liquid waste (HLLW). Ru tetra-oxide, RuO₄, has high vapor pressure, which is high enough to be run off from its solution even at room temperature. Electrochemical oxidation method, to oxidize nitrosyl ruthenium to the tetra-oxide and then to remove ruthenium from liquid phase to gas phase, was studied to separate Ru from the HLLW. The advantage of this method requires neither additional reagents nor adjustment its valency before the oxidation and disadvantage is necessity of long time for oxidation. In order to improve oxidation rate, we carried out the experiments to clarify the effects following fundamental conditions to the electrochemical oxidation, which are (a) electrolyte temperature, (b) presence of promoter elements, (c) evaporation or reflux of condensed phase, and (d) using or not using of diaphragm at counter electrode. We found the fast oxidation conditions as follows: (1) higher temperature; 95°C, (2) Ce coexistence; 3000 ppm; and (3) usage of a diaphragm for counter electrode. However, evaporation or reflux conditions did not directly affect the electrochemical oxidation efficiency.     

Separation of actinoids from HLW by thiacalix[4]arene compound impregnated silica ion-exchanger

If trivalent actinoids such as Am can be separated from high-level radioactive liquid waste (HLLW) generated by the PUREX process, long-term heat and radiation hazards can be significantly reduced in the disposal of this waste. This paper reports the effective separation of Am through the use of a chelating ion-exchange method that uses an octylphenyl-N,N-diisobutylcarbomoyl phosphine oxide (CMPO) or thiacalix[4]arene compound impregnated silica ion-exchanger. The separation of Cs and Sr from HLLW can be achieved using a CMPO impregnated silica ion-exchanger (CMPO-exchanger). Actinoids and lanthanoids can then be eluted from the CMPO-exchanger, with the resulting solution then treated to separate the actinoids using a thiacalix[4]arene compound impregnated silica ion-exchanger. Thiacalix[4]arene compound impregnated silica ion-exchangers have been shown to effectively separate actinoids from lanthanoids in a weak-acid solution. The influence of gamma ray irradiation on adsorption is also investigated. The adsorption of Am remains high even after irradiation of the ion-exchanger.     

Crystalline phases in a devitrified simulated high-level waste glass containing the elements of the platinum group

A borosilicate glass containing 20 wt% simulated high-level waste oxides was subjected to heat treatment at 700°C for 1000 h. Seven crystalline phases were newly formed by the treatment in addition to two phases, (Ru, Rh)O₂ and (Pd, Rh, Te), which had already existed in the as-prepared simulated high-level waste glass. Among the new seven phases, five phases were certainly identified to be (RE)BSiO₅, CeO₂, SiO₂, (RE)PO₄ and (Sr, Ba, RE)MoO₄. Of two unidentified phases, one was rich in silicon, chromium and rare earth elements (RE), and the other was rich in nickel and chromium.The crystalline phases of the elements of the platinum group facilitated the occurrence of the other phases and suppressed crystal growth.     

Separation of rare metal fission products in radioactive wastes in new directions of their utilization

An electrolytic extraction method has been studied to separate fission products (Ru, Rh, Pd, Tc, Se, Te, etc.) from the nuclear spent fuel. Yet they are rare metal fission products (RMFP), most are long-lived (LLFP; Pd, Tc, Se, Te). In the applied electrochemical separation process, Pd²⁺ cation itself would not only be easily deposited from various nitric acid solutions, but enhance also the deposition of RuNO³⁺ and ReO₄⁻ by acting as a catalyst (as Pd_adatom). Such Catalytic Electrolytic Extraction (CEE) method was found to be applicable in the case of TcO₄⁻ deposition. The quaternary-, Pd-Ru-Rh-Re, deposit Pt electrodes showed the highest cathodic current corresponding to the hydrogen evolution reaction, ca. twice superior to that of the Pt electrode in artificial sea water as well as in alkaline solution. Adsorption / desorption behavior of ¹⁰⁶Ru, etc. on the tertiary pyridine resin, being set as a pre-separation step to the CEE process, was newly confirmed by the reprocessing experiments with a highly irradiated fast reactor MOX fuel. The promising utilization of recovered RMFP will be a “FP-catalyst” for electrolytic hydrogen production, where they would be circulating material to bridge nuclear and hydrogen energy systems.     

AR-01树脂对铼的吸附和解吸行为

99 Tc是核电站运行过程中产生的裂变产物之一,是一种具有较高裂变产额的长寿命放射性核素,易在环境中迁移,因此高放废液中锝的分离对于核废料后处理具有重要意义。以铼模拟锝,通过静态实验和动态实验研究了AR-01树脂对铼的吸附和解吸行为,主要考察了温度、酸度、时间等因素对AR-01树脂吸附性能的影响,并通过还原解吸法探讨解吸条件。结果表明:AR-01树脂吸附铼的反应为放热反应,酸度提高,吸附效率降低,在流速1.0mL/min下,树脂的动态饱和吸附容量达到105mg/g,解吸剂中加入还原剂SnCl2提高了解吸效率,树脂具备很好的重复利用性。以上结果为AR-01树脂应用于高放废液中锝的吸附分离提供了参考。     

Experiments on the leak path factor of ruthenium volatilized from high-level liquid waste tanks in a reprocessing plant in case of the boiling and drying accident

In case of the boiling and drying accident of a high-level liquid waste (HLLW) tank, a large amount of ruthenium (Ru) will be volatilized. In order to suppress the release of radioactive materials to the environment, the vapor may be led to the neighboring cells in which it will contact with the cell walls to be partially condensed. To understand the behavior of Ru in this situation, we have prepared an experimental apparatus. It consists of a small tank in which 60 mL simulated HLLW is heated to dryness, a 9.6 L stainless steel box which mimics the neighboring cell accepting the vapor from the small tank, and a condenser where the vapor coming out from the box is cooled to collect the condensate. The results show that more than 99% of the volatilized Ru is removed from the vapor in the box if its temperature is below about 120 °C.     

Characterization of alloy particles extracted from spent nuclear fuel

We characterized, for the first time, submicro- and nanosized fission product-alloy particles that were extracted nondestructively from spent nuclear fuel, in terms of noble metal (Mo–Ru–Tc–Rh–Pd–Te) composition, atomic level homogeneity and lattice parameters. The evidences obtained in this work contribute to an improved understanding of the redox chemistry of radionuclides in nuclear waste repository environments and, in particular, of the catalytic properties of these unique metal alloy particles.     

Removal of Platinum Group Metals Contained in Molten Glass Using Copper

Removal of platinum group metals (PGMs) such as Pd, Ru, and RuO₂ from molten glass by using various amounts of liquid Cu was done as a basic study on a new vitrification process for a high-level radioactive waste. We prepared two types of borosilicate glasses containing PGMs and Cu, respectively. These glasses were mixed together and heated at 1,473 K for 4 h in Ar atmosphere. More than 95% of Pd were removed as a spherical metal button composed of Pd-Cu alloy when Cu was added in an amount 0.5 times the weight of Pd. Nearly 95% of Ru was also removed as a spherical button with 2.5–5 times as much Cu addition as Ru in weight. Ruthenium oxide was reduced to metallic Ru by a reaction with Cu in the molten glass. The removal fraction was increased by increasing the amount of Cu and reached 63% when Cu addition was 7.5 times as much as RuO₂ in weight. By addition of Si as a reducing agent, nearly 90% of Pd and Ru were removed with Cu and Si metal composites even under O₂:Ar = 20:80 (v/v) condition.     

A new back-end cycle strategy for enhancing separation, transmutation and utilization of materials (Adv.-ORIENT cycle)

To minimize the ecological burden originating in nuclear fuel recycling, a new R&D strategy, the Adv.-ORIENT (Advanced Optimization by Recycling Instructive Elements) cycle was set forth. In this context, mutual separation of f-elements, such as minor actinide (MA)/lanthanide (Ln) and Am/Cm, are essential to enhance the MA (particularly ²⁴¹Am) burning. Isotope separation before transmutation is also inevitably required in the case of some long-lived fission products (LLFPs) like ¹²⁶Sn, ¹³⁵Cs, etc. The separation and utilization of rare metal fission products (RMFPs: Ru, Rh, Pd, Tc, Se, Te, etc.) are offering a new direction in the partitioning and transmutation (P&T) field. ⁹⁹Tc and ¹⁰⁶Ru, well-known interfering nuclides in reprocessing, should be removed prior to the actinide stream. Separation of exothermic nuclides ⁹⁰Sr, ¹³⁷Cs as well as MA will significantly help to mitigate the repository tasks.

A key separation tool is ion exchange chromatography (IXC) by a tertiary pyridine resin having soft donor nitrogen atoms. This method has provided individual recovery of pure Am and Cm products with a Pu/U/Np fraction from irradiated fuel in just a 3-step separation. A catalytic electrolytic extraction (CEE) method by Pd_adatom has been employed to separate, purify and fabricate RMFP catalysts. Differently functioned ion exchangers, e.g., ammonium molybdophosphate (AMP), have been investigated for the separation of Cs⁺. Theoretical and laboratory studies on the isotope separation of LLFPs were begun for ⁷⁹Se, ¹²⁶Sn and ¹³⁵Cs.     

Evaluation study on properties of a macroporous silica-based CMPO extraction resin to be used forminor actinides separation from high level liquid waste

Basic properties of a silica-based octyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide (CMPO) extraction resin (CMPO/SiO2-P) was investigated.Adsorption behavior for some rare earth elements (RE) which are constituents of high level liquid waste (HLLW) and the long-term stability of the extraction resin in nitric acid solution were examined.The CMPO extraction resin was significantly stable in 3 mol·L?1 HNO3 solution at 50oC.Furthermore,the RE(III) were efficiently separated from non-adsorptive fission product (FP) elements such as Sr(II) in a column experiment using a highly nitric acid solution.The separation behaviors of the elements are considered to result from the difference in their adsorption and elution selectivity based on the complex formation with CMPO.There was no strong dependency of RE(III) separation efficiency on feed solution flow rate.Only from the perspectives of the acid-resistant behavior of CMPO extraction resin and the elution kinetics for the metal ions with the extraction resin,the CMPO extraction resin can be used in the modified MAREC process for HLLW partitioning.     

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.     

拉曼光谱法直接测定模拟高放废液中硫酸根

对影响拉曼光谱测量硫酸根的主要因素,如积分时间、叠加次数、激光功率等进行了研究,确定了最佳测量条件,建立了拉曼光谱测量模拟高放废液中硫酸根的分析方法。该方法利用拉曼光谱仪测定硫酸根与水拉曼峰的相对强度,从而得到硫酸根含量。结果表明:硫酸根质量浓度线性范围为1~15g/L,线性相关系数为0.999 6,相对误差小于±5%,检测限为0.05g/L。此方法简单、快速、无需消耗任何试剂,有望用于某高放废液样品中硫酸根的检测。