Interface Characteristics in Settler of Centrifugal Extractor

Centrifugal extractor has the shortest contact time and it is to be used for FBR fuel reprocessing for the future. The two liquid phases are separated by centrifugal force in the settler and the light liquid flows out through the inside weir and the heavy liquid through the outside one. And the interface has to be hold to keep stable operation. However, since the observation of the interface was very difficult, the behavior of the interface in the settler was not clearly understood before. In this study, in order to observe the correct diameter of the interface, the experimental equipment is made of transparent plastic and the diameter is determined by taking photos, lighting a flash. As a result, the stability of the interface is examined for various flow rates and rotation velocities. From the data, it is shown that the pressure drops at 1,800 rpm is proportional to the flow rate at organic weir but is proportional to the power of 0.64 of the flow rate at aqueous weir.     

Development of COMPAS,Computer Aided Process Flowsheet Design and Analysis System of Nuclear Fuel Reprocessing

A computer aided process flowsheet design and analysis system, COMPAS has been developed in order to carry out the flowsheet calculation on the process flow diagram of nuclear fuel reprocessing. All of equipments, such as dissolver, mixer-settler, and so on, in the process flowsheet diagram are graphically visualized as icon on a bitmap display of UNIX workstation. Drawing of a flowsheet can be carried out easily by the mouse operation. Not only a published numerical simulation code but also a user's original one can be used on the COMPAS. Specifications of the equipment and the concentration of components in the stream displayed as tables can be edited by a computer user. Results of calculation can be also displayed graphically. Two examples show that the COMPAS is applicable to decide operating conditions of Purex process and to analyze extraction behavior in a mixer-settler extractor.     

后处理工艺Purex流程计算机模拟研究现状及展望

利用与Purex流程相关的基础数据,开展Purex流程计算机模拟研究并形成模拟程序,能够开展工艺条件分析和工艺优化工作,具有重要的应用价值。国外对于此类研究开展的较早,在分配比模型研究上形成了以Richardson模型为代表的半理论模型;混合澄清槽和脉冲萃取柱的计算机模拟也分别在全混模型和扩散模型的基础上开展了大量的研究工作,形成了较多的模拟程序。我国开展此类研究稍晚,仅在分配比模型和混合澄清槽模拟方面开展了部分研究工作,与国外存在较大的差距。     

Development of hybrid reprocessing technology based on solvent extraction and pyro-chemical electrolysis

Toshiba has been proposing a new fuel cycle concept for the transition period from Light Water Reactors (LWRs) to Fast Reactors (FRs). This concept involves a more valuable process for LWR spent-fuel reprocessing than the conventional process and improved proliferation resistance. We have been developing a new technology, the Toshiba Hybrid Reprocessing Process, based on solvent extraction and pyro-chemical electrolysis, for spent fuel reprocessing for the transition period from LWRs to FRs. The Toshiba Hybrid Reprocessing Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high-purity uranium (U) and the pyro-chemical process in molten salts for recovery of impure plutonium with minor actinides (Pu + MA). High-purity U is used for LWR fuel, and impure Pu + MA is used for metallic FR fuel. Valence control by the electrolysis and solvent extraction tests using LWR spent fuel and oxalate precipitation tests were carried out to confirm the feasibility of the Toshiba Hybrid Reprocessing Process. A consecutive processing equipment for the solvent extraction process and a bench-scale apparatus for the pyro-chemical process were manufactured. The consecutive processing equipment consists of a flow type electrolytic cell and a centrifugal extractor. The test revealed that U of 99.99% of purity was recovered. The bench-scale apparatus consists of a reactor for oxalate precipitation, a solid-liquid separator in which nitric acid with fission products and precipitation are separated, and a drying equipment in which the precipitation is dry. Precipitation test with neodymium (Nd) which is simulated as Pu + MA in nitric acid was carried out. It was confirmed that precipitation ratio of Nd was more than 99.9% and that moisture ratio of the precipitation was less than 10%. The results suggested that U recovery of LWR spent fuel was 99.99% with the consecutive processing equipment and Pu + MA recovery was more than 99.9% with the bench-scale apparatus. The Toshiba Hybrid Reprocessing Process could recover high-purity U used for LWR fuel, and impure Pu + MA used for metallic FR fuel.     

Recent advances of annular centrifugal extractor for hot test of nuclear waste partitioning process

Advances are being made in the design of the annular centrifugal extractor fornuclear fuel reprocessing extraction process studies.The extractors have been built and tested.Twelve stages of this extractor and 50 stages are used toimplement the TRPO process for the cleanup ofcommercial and defense nuclear waste liquids,respectively.Following advances are available:(1) simple way of assembly and disassembly between rotor part and housing part of extractor,ease of manipulator operation;(2)automatic sampling from housing of extractor in hot cell;(3) compact multi-stage housing system;(4) easy interstage link;(5) computer data acquisition and monitoring system of speed.     

核燃料水法后处理现状和展望

本文评论了世界各国乏燃料后处理技术。后处理能力在目前和不久将来不能满足核能的发展需要,它在燃料循环中占有重要的地位。普雷克斯流程不仅对于轻水堆燃料,而且对快中子增殖堆燃料后处理然是一种主要流程。近期后处理研究和发展的重点在于使流程最佳化,并引入新技术,尤其是U(Ⅳ),电解氧化还原,硝酸羟胺还原和亚硝气氧化等无盐过程。当处理高燃耗的乏燃料时,应采取特别措施,以避免造成溶剂的严重辐解和钚的临界问题。氚在流程中必须控制,并限定在一定区域,使其废液体积尽量减小。     

Hydrodynamic and mass transfer studies of 125 mm centrifugal extractor with 30% TBP/nitric acid system

The Indian nuclear power programme was conceived with a three-stage structure in order to utilize the resources optimally with at most importance to fuel reprocessing for closing the fuel cycle. The first stage of Indian nuclear power programme is based on natural uranium fuelled pressurized heavy water reactors to produce the plutonium (Pu) feed for the second stage. The second stage consists of plutonium fuelled fast breeder reactors to produce U-233 from thorium. The third stage envisages development and deployment of U-233 fuelled reactors. In the fuel cycle operations, solvent extraction is a major step for the recovery of uranium and plutonium. Centrifugal extractors plays a vital role in solvent extraction due to their compact size and high throughput. In the present work, the experimental studies for the hydraulic performance were reported for a single stage annular centrifugal contactor of ϕ125 mm rotor for two-phase flow. Maximum throughput with entrainment less than 1% of one phase to the other, of the centrifugal contactor was measured with the A/O ratio 1 to 6 for the three different bottom vane heights of 6, 8 and 10 mm and also for three different annular gaps of 12, 15 and 18 mm. The centrifugal extractor was operated at different speeds ranging from 1200 to 2200 rpm. In addition, mass transfer performance of the same unit was evaluated with 30% TBP/nitric acid biphasic systems.     

A new concept of nuclear fuel reprocessing by applying ion-exchange technology

The development of advanced technology for the spent nuclear fuel reprocessing should be achieved not only considering cost, non proliferation and reduction of radioactive wastes but also corresponding to both spent nuclear fuels of LWR and FBR.

We have proposed an ion exchange process for reprocessing using a new type ion exchanger developed to chemical method of U enrichment technology. This process possess possibility of a sharp cut in cost, since this ion exchanger is characterized by rapid adsorption-desorption rate dominating the treatment rate.

From the basic experimental results, this reprocessing process has been constructed by 3 ion exchanger columns which consist of a main separation column, the uranium-refining column and the plutonium-refining column.

Comparing ion exchange process with the conventional Purex process, this ion exchange process has many advantages such as the decrease in the number and size separation equipment, solvent-spent free and alkaline-liquid-spent free. With these advantages, this process is estimated that the construction cost of reprocessing process is greatly reduced comparing to the conventional process.     

Studies on physics parameters of metal (U–Pu–Zr) fuelled FBR cores

For faster growth of nuclear power in India, it is essential to shift to the use of metal-fuels in fast breeder reactors (FBR), which gives a higher breeding ratio (BR) and lower doubling time (DT). Also, future commercialization of the FBR fuel cycle necessitates the use of metallic fuel along with the pyro-process recycling, which can be less costly than oxide fuel reprocessing. Two-dimensional diffusion calculations have been performed to investigate the various physics parameters of metal (U–Pu–Zr) fuelled FBR cores as a function of reactor parameters like reactor power, smear density, zirconium content in the fuel and the number of rows in radial blankets. A 1000 MWe fast reactor with U–Pu fuel (i.e. metal-fuel with no zirconium – which is a theoretical possibility now, due to the lack of irradiation experience) can attain a breeding ratio of 1.61 and a reactor fuel doubling time of 6.6 yrs. Two methods to reduce the sodium void reactivity, which is high and positive in metal-fuelled FBR cores, are suggested.     

Single-Stage Extraction Test with Continuous Flow of Molten LiCl-KCl Salt and Liquid Cd for Pyro-Reprocessing of Metal FBR Fuel

A single-stage extraction test with the continuous flow of both molten salt and liquid Cd was carried out as a preliminary step of the development of a multistage countercurrent extraction process for the pyrometallurgical reprocessing of metal FBR fuel. Ce, Gd, and Y were selected as the substitutes for U, other actinides, and rare-earth elements, respectively. More than 97% of Ce could be recovered in one of the experiments, in which 97% of Gd and 63% of Y were also recovered. This meant that high recovery yield could be obtained in the system, but mutual separation was not sufficient when the high recovery yield was required in a single-stage extraction. In another experiment in which the recovery yields of Ce, Gd, and Y were about 87, 70, and 8%, respectively, the separation factors of Gd and Y against Ce were 0.27 and 0.0092, respectively, which were about the same as their equilibrium values. The result showed that higher separation efficiency was achieved under the condition of relatively lower recovery yield. The mass transfer coefficient of rare-earth elements in the salt was estimated to be 0.14 cm/s, which was much larger than that in the previous experiments without a baffle plate in the extractor.     

Extractor Type in Liquid Metal-Molten Salt Extraction System

Two factors, which are important for selecting the extractor type suitable for liquid metal-molten salt system were studied, i.e. the formation and the coalescence processes of liquid metal drop. The drop formation process for liquid metal dispersion in the continuous phase is predictable from semi-empirical correlation reported on aqueous solution-organic solvent systems. The height of droplet bed being accumulated on drop-settling portion is predictable from the coalescence time of single drop on a fiat metal interface. The coalescence of metal drop on a clean interface was very fast. The extractor type of liquid metal dispersion in molten salt is considered to be suitable for the pyrochemical extraction process.     

Purification of uranium products in crystallization system for nuclear fuel reprocessing

Uranium crystallization system has been developed to establish an advanced aqueous reprocessing for fast breeder reactor (FBR) fuel cycle. In the crystallization system, most part of uranium in dissolved solution of spent FBR-MOX fuels is separated as uranyl nitrate hexahydrate (UNH) crystals by a cooling operation. The targets of U yield and decontamination factor (DF) on the crystallization system are decided from FBR cycle performance and plutonium enrichment management. The DF is lowered by involving liquid and solid impurities on and in the UNH crystals during crystallization. In order to achieve the DF performance (more than 100), we discuss the purification technology of UNH crystals using a Kureha Crystal Purifier (KCP). Results show that more than 90% of uranium in the feed crystals could be recovered as the purified crystals in all test conditions, and the DFs of solid and liquid impurities on the purified UNH crystals are more than 100 under longer residence time of crystals in the column of KCP device. The purification mechanism is mainly due to the repetition of sweating and recrystallization in the column under controlled temperature.     

Radiation Damage of Organic Extractant in Partitioning of High-Level Liquid Waste, (II)

Radiation dose absorbed in the organic extractant was estimated for the counter- current extraction with mixer-settler in the partitioning of the high-level liquid waste of fuel reprocessing. The radiation dose due to the absorption of α and β-particles was calculated to be 10, 5.0 and 0.42 Wh/l for 150 days, 1 year and 10 years after fuel discharge from the reactor, respectively. About 70–90% of the total dose was due to the absorption of β-particles. These values were discussed in comparison with the published data for the fuel reprocessing. On the basis of the calculated data, the radiolytic effects upon the extraction of strontium, lanthanoids and transplutonium elements in partitioning were estimated be sufficiently small.     

铀钚萃取洗涤-共反萃工艺的离心萃取器台架实验研究

由于快堆MOX乏燃料放射性强,需要缩短停留时间以降低溶剂辐解,本工作以离心萃取器为萃取设备,在短停留时间下进行了快堆MOX乏燃料后处理铀钚萃取洗涤-共反萃工艺研究。研究结果显示,该工艺在单级停留时间约20s时具有良好的铀钚收率,萃取洗涤过程中铀和钚收率均大于99.99%,共反萃过程中铀和钚收率分别为99.99%和99.94%;同时能有效防止第三相的形成,避免钚的聚合沉淀。     

乏燃料后处理中的辐射化学问题 Ⅰ.萃取剂的辐射稳定性

由于乏燃料具有强辐射性的特点,辐射化学伴随在乏燃料后处理每个过程中。尽管α和γ等电离辐射对于萃取剂本身的直接效应一般不大,但是它们通过与水相和油相中的溶剂相互作用产生的自由基,一方面可以攻击萃取剂的配位基团,另一方面溶剂辐解产生的活性粒种可能与金属离子反应改变其氧化态,从而降低其萃取效率或分配比。水相中硝酸辐解产生的亚硝酸对于金属离子的氧化态会产生重要影响,产生的自由基如•NO₃等也会与萃取剂反应使其劣化。在先进核能系统中,随着燃耗的提高,放射性更强,而且用于溶解乏燃料的硝酸浓度也增高,因而,对于先进核燃料循环中的辐射化学研究既是良好机遇也是重大挑战。本文旨在对近十年来国内外乏燃料后处理（溶剂萃取）方面有关辐射化学研究特别是新型萃取剂的辐射稳定性等进行综述与讨论。     

后处理厂1A柱有机相着火事故源项估算

我国对于后处理工业的需求随着核电事业的迅猛发展变得愈发强烈,为了满足后处理工业安全发展必不可少的核应急需求,为核应急工况下后处理厂的核应急响应与决策支持提供依据。针对后处理厂1A柱有机相着火事故这一基准事故,结合实际工艺流程及监测手段,选取了核应急工况下的可获得参数(有机溶剂泄漏质量等)作为输入,在有机相燃烧速率经验公式基础上,结合后处理的工艺特点,引入少量修正建立了后处理厂1A柱有机相着火事故源项估算模型,并使用FORTRAN编程语言开发了相应软件。数值验证结果表明,该估算模型可以满足后处理厂1A柱有机相着火事故的核应急需求。     

Description of Fuel Integrity Project methodology principles

Abstract

TN International (TNI) and International Nuclear Services (INS) started the Fuel Integrity Project (FIP) in early 2000s, with the goal of developing a methodology to evaluate, as a safety requirement, the nature and the extent of fuel assemblies (FA) damage during accident drops of a packaging. From TNI initial knowledge acquired from fresh FA behaviour during drop tests, a mechanical tests programme, including testing on fresh and used fuel rod samples, has been planned by both companies and executed by INS. Tests results analysis has led to the elaboration of FIP methodology by TNI. Experimental knowledge was collected from the testing programme and the main mechanical phenomena arising from a drop have been identified and quantified. As a result, the FIP methodology, structured in flow charts, gives guidelines to study the effects of a lateral or axial drop of a packaging loaded with fresh or used FA of pressurised water reactor or boiling water reactor types. The flow charts of methods have the same philosophy: several conservative mechanical evaluations based on direct calculations or dimensionless comparisons with appropriate reference tests permit determination of FA damage which gradually increases with acceleration. First, elastic models distinguish the null or slight damage cases; then, plastic models identify cases with extreme FA damage that lead to unacceptable criticality hypotheses; finally, other plastic models quantify the extent of fuel rod deformations in moderate FA damage cases. Fuel Integrity Project methodology application to a given case leads to the following output, used as criticality hypotheses for the safety analysis: existence or not of fuel rods rupture, their number, their location, the associated amount of released fuel material, and the extent of fuel rods array deformation and sliding. All the knowledge arising from the FIP is compiled in the ‘Technical guide’, which presents extensively the methodology and all background experimental data. The methodology is applicable to fresh and used FA, assuming that the brittle fracture risk is analysed otherwise.     

离心萃取级联分离锂同位素的数学模型

萃取法分离锂同位素有望替代汞齐法消除汞害,但需多级萃取才能获得高丰度同位素,采用离心萃取机替代萃取澄清槽形成萃取级联系统可提升分离效率。基于萃取法分离锂同位素、离心萃取分离原理和级联理论,借鉴气体离心级联分离同位素的方法,引入分流比概念,建立了离心萃取级联分离锂同位素单级、多级的数学模型和级联的平衡时间模型,对离心萃取级联分离锂同位素进行计算分析。离心萃取级联是一种类似全回流矩形级联形式,取料量对级联级数有着很大的影响,级联存在最大取料丰度限制,级联平衡时间受到目标丰度和离心萃取机级停留时间(处理能力)影响,采用多步法级联可有效减少平衡时间。该数学模型可指导工艺的设计,为下一步的产业化应用提供理论依据。     

Φ10mm离心萃取器水力学性能研究

为将Φ10mm离心萃取器用于Purex流程钚纯化循环反萃段的实验研究中,当水相和有机相流比为1∶4,离心萃取器的重相堰直径分别为6.4、6.6和6.8mm时,考察了两相出口料液的夹带情况以及环隙和转筒内液体体积随转速的变化等水力学性能。研究表明,当两相总流量小于9.0mL/min、转速大于4000r/min时,离心萃取器处于稳定可操作区间。此时两相出口料液均不夹带,两相混合区内液体量约为0.7mL,转筒内液体量约为2.2mL。结合总流量可进一步计算得出两相接触时间。     

乏燃料后处理中的辐射化学问题Ⅱ.水溶液和稀释剂的辐射分解

由于乏燃料具有强辐射性的特点,辐射化学伴随在乏燃料后处理每个过程中。尽管α和γ等电离辐射对于萃取剂本身的直接效应一般不大,但是它们通过与水相和油相中的溶剂相互作用产生的自由基,一方面可以攻击萃取剂的配位基团,另一方面溶剂辐解产生的活性粒种可能与金属离子反应改变其氧化态,从而降低其萃取效率或分配比。水相中硝酸辐解产生的亚硝酸对于金属离子的氧化态会产生重要影响,产生的自由基如·NO_3等也会与萃取剂反应使其劣化。在先进核能系统中,随着燃耗的提高,放射性更强,而且用于溶解乏燃料的硝酸浓度也增高,因而,对于先进核燃料循环中的辐射化学研究既是良好机遇也是重大挑战。本文重点对近十年来国内外在乏燃料后处理(溶剂萃取)方面有关辐射化学研究,特别是硝酸的辐射分解、锕系水溶液的辐射化学、稀释剂的辐射分解等方面进行综述与讨论。