Behavior of Ruthenium in Fluoride-Volatility Processes, (VI)

Hydrogen gas evolution from water dispersing nanoparticles induced by ⁶⁰Co γ-ray irradiation was studied. Nanoparticles of TiO₂ and Al_2O3 with average sizes of 7–33 nm supplied from several suppliers were examined. It was indicated that reactions enhancing the hydrogen evolution proceed on particle's surface. It was implied that the yield depends on size of agglomerated particle regardless of their primary particle size and chemical species. Reactions that enhance the hydrogen yields were discussed, and radiolysis process was concluded dominant in the total enhancement mechanism.     

Experimental study on elemental behaviors in fluorination of nuclear spent fuel with flame reactor

Our proposed spent nuclear fuel reprocessing technology named FLUOREX is a hybrid system based on reprocessing technologies of fluorination and solvent extraction for light water reactor fuel. In the current research, we experimentally clarified solid–gas transfer behaviours of the fluorides in the FLUOREX process and identified the volatile and non-volatile compounds in the fluorination. We carried out a fluorination experiment for simulated spent nuclear fuel and solid separation from the UF₆ gas stream. The distribution ratios of fission product elements in the experimental apparatus were evaluated. Molybdenum, Te, Nb, and Ru were volatilized by fluorination and they accompanied the UF₆ gas. However, 22.9% of the Ru and 3.4% of the Nb were retained as solids in the experimental apparatus, contrary to the fact that their partial pressures in the experiment were lower than their vapor pressures. Rubidium, Sr, Zr, Ce, and Nd were completely recovered as solid fluorides, and these results agreed with the prediction based on boiling points of their fluorides. Antimony was completely recovered as a solid; nevertheless, the boiling point of antimony pentafluoride was lower than the process temperature, and that was attributed to the formation of a non-volatile antimony oxyfluoride.     

Solubility of Tributylphosphate in Plutonium Nitrate Solution and Highly Radioactive Liquid Waste Solution

Abstract

The solubility of tri-n-butylphosphate (TBP) in aqueous solutions of plutonium nitrate (PuN) and in highly radioactive liquid waste (HRLW) of PUREX nuclear fuel reprocessing was investigated. By an empirical formula the solubility of TBP in PuN solutions was described in the range of 0–0. 1 M Pu and 1–8M HNO₃ concentrations. The following items were elucidated:

(1) The logarithm of TBP solubility (S) in the solution of interest varies inversely in proportion to the concentration of Pu(IV) in the range of 0–0.1M PU(IV) at a constant concentration of HNO₃, indicating that Pu(IV) simply behaves as an electrolyte for the salting-out of TBP. Log S subsequently levels off with increasing Pu concentration, which would be due to a change in the principal dissolution form of TBP having an interaction with Pu (IV).

(2) The variation in S in PuN solutions (0–0.1M PU) with nitric acid concentration shows almost the same tendency as that in HNO₃ solution.

(3) A dependency of S on fission product metal ions in HNO₃ for HRLW similar to that for PuN was observed.

(4) The logarithm of the ratio of TBP solubility in water to that in solution of interest was nearly proportional to l/T for HRLW solution or for low concentration of PuN solution. That deviates from the linear relation at high temperature when the concentration of PuN is increased, which can be explained by the change in ionic form of Pu.     

正己烷与正十二烷在硝酸体系中的化学行为

通过红外光谱、气相色谱-质谱联用和核磁共振碳谱等分析方法确定了正己烷和正十二烷直链烷烃在硝酸体系中的反应产物,用气相色谱法定量分析了温度、硝酸浓度和反应时间等因素对正己烷和正十二烷硝化产物的影响,利用定制的高压反应釜考察了正己烷、正十二烷在硝酸体系及温度为150℃情况下的压力变化情况。研究结果表明:高温下正己烷和正十二烷在硝酸体系中的反应产物分别为硝基己烷异构体混合物(2-硝基己烷和3-硝基己烷)和硝基十二烷异构体混合物(2-硝基十二烷、3-硝基十二烷、4-硝基十二烷、5-硝基十二烷和6-硝基十二烷),且温度及硝酸浓度对硝化产物影响较大,而反应时间则影响较小。温度为150℃时,正己烷在硝酸体系中产生的最大压力为3.44 MPa,远大于正十二烷体系下的1.58 MPa,且随着硝酸比例的增加反应体系产生的压力及反应程度均有所增加。     

Measurement of homogeneity of U and Pu isotopes in PWR spent fuel powder mixture

We present a feasibility study of the homogenization of pressurized water reactor spent nuclear fuel (SNF) powder through a mechanical mixing process. Because burn-up of the SNF depends on the position in the SNF assembly, concentrations of uranium, plutonium, and fission products are distributed differently according to the burn-up profile. The heterogeneity of the material elements affects the selection of a representative sample for quantitative analysis. Homogenization process improvement to reduce the sampling error is thus required to precisely determine the amount of uranium, plutonium, and fission products in the SNF. In this study, fine powders (<70 μm) extracted from one SNF rod were mixed, and the degree of homogenization was determined as a function of the mixing time indicating the relative standard deviations of the ¹³⁴Cs/¹³⁷Cs, Pu, and U isotope ratio measurements.     

The Application of an Advanced Ion Exchange Process to Reprocessing Spent Nuclear Fuels, (I)

In order to develop an advanced ion exchange process for the reprocessing of spent nuclear fuels, a novel anion exchanger, AR-01 with the resin embedded in porous silica beads and benzimidazoles as functional groups has been manufactured. Adsorption behavior of various fission product elements (FPs) and uranium in nitric acid medium were investigated experimentally using this anion exchanger. Separation performance of FPs from U(VI) in simulated spent fuel solutions was demonstrated by column chromatography utilizing dilute HNO₃ and thiourea as eluents.

Most FPs such as Cs(I), Sr(II), Mo(VI), Rh(III) and trivalent rare earths showed negligibly slight adsorption and could be separated from U(VI) satisfactorily. Cerium(IV) was strongly adsorbed, but was gradually reduced to non-adsorptive Ce(III) by the anion exchanger. Zirconium(IV) presented weak adsorption and its a part mixed with U(VI) in the column experiments. Ruthenium(III) exhibited quite strong adsorption in a broad HNO₃ concentration range as the form of anionic nitrosylnitrato-complexes, its most amount mixed with U(VI). Palladium(II) showed significantly strong adsorption probably due to complexes formation with the anion exchanger. The adsorbed Pd(II) was effectively eluted out by thiourea and separated from U(VI) and other FPs completely.     

Ion Exchange Technology in Spent Fuel Reprocessing

Solvent extraction is the major unit operation employed in spent nuclear fuel reprocessing. The operation yields three streams; fission product waste, uranium product and plutonium product. Ion exchange is primarily used in reprocessing as a tail-end method to concentrate and isolate the plutonium product stream. This review will describe the details of plutonium recovery and purification by both cation- and anion-exchange processing. A brief overview of miscellaneous uses of ion-exchange employed in reprocessing will also be given.     

溶液中钚α活度的直接测量研究

选取YAP∶Ce闪烁体作为仪器核心部件,建立了接触式测量溶液中微量、痕量钚α活度的分析方法。结果表明:水相钚溶液浓度在5.20×10-5~1.30×10-3 g/L范围内线性良好(R2=0.987 7),定量检测下限为5.20×10-5 g/L;有机相钚溶液浓度在2.27×10-5~1.13×10-3 g/L范围内线性良好(R2=0.992 3),定量检测下限为2.27×10-5 g/L。本分析方法有别于传统的α计数法,过程无需制源,操作简单、方便,有望为后处理工艺过程中微量和痕量钚的在线或实验室分析提供一种新的途径。     

Dissolution of ThO2Based Oxides in Nitric Acid Solutions at Elevated Temperatures

Highly-dense spherical particles of thorium-based oxides, ThO₂ and (Th, U)O₂, prepared by the sol-gel method were subjected to dissolution with nitric acid containing 0–0.05 mol/l NaF at high temperatures above 120°C. The dissolution rate depended upon temperature, fluoride concentration and UO₂ content. High-temperature in the range of 120–200°C enhanced the dissolution of the ThO₂-based fuels. At low temperatures and/or low U0₂ concentrations, insoluble tetrafluoride precipitates were formed on the particle surfaces and they resulted in the decrease of the dissolution rates. In the present study, the apparent activation energies for the high-temperature dissolution were obtained.     

Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle

Electrochemical methods for the separation of fission products from fission material in molten fluoride salt media have been studied in the context of their application within the framework of the developed Molten Salt Reactor fuel cycle. The separation possibilities of selected actinides (U, Th) and lanthanides (Nd, Eu, Gd) in molten eutectic LiF-NaF-KF at 530°C were evaluated by means of cyclic voltammetry. The applicability of different electrochemical techniques is discussed with reference to the new results from this study, and a basic flow sheet for the Molten Salt Reactor fuel cycle is outlined.     

Dissolution behavior of (U,Zr)O2-based simulated fuel debris in nitric acid

To explore the possibility of dissolving fuel debris into nitric acid as a potential pre-treatment for waste treatment in which the U and Pu are removed from the inventory, dissolution tests of U_1?xZr_xO₂ and (U,Pu)_1?xZr_xO₂ were carried out in 6 M HNO₃ at 353 K. At the end of the dissolution test (after 4 h), the ratio of dissolved uranium decreased with an increase in the Zr contents, x. While the dissolution of U-rich samples was congruent, a preferential leaching of U was observed with Zr-rich samples. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). The IDR decreased from 10^?5 down to 10^?10 mol cm^?2 min^?1 as x increased from 0 to 0.95. From these findings, dissolution with HNO₃ is expected to be only applicable in U-rich part of fuel debris (x < 0.3) if the dissolution in 6 M HNO₃ at 353 K is assumed. Application of complexing acids, such as mixture of HNO₃ and HF, should be considered to increase the dissolution rate of the Zr-rich part.     

Uranium Crystallization Test with Dissolver Solution of Irradiated Fuel

The crystallization process has been developed as a part of the advanced aqueous process, NEXT (New Extraction System for TRU recovery) for fast reactor (FR) cycle. In this process, a large part of U is separated from dissolver solution by crystallization as UO₂(NO₃)₂.6H₂O. The U crystallization test was carried out with real dissolver solution of irradiated FR fuel to investigate the influence of cooling rate on the crystal size and the behavior of fission product (FP) compared with that of Pu(IV). In regard to the influence of the cooling rate, it was confirmed that the crystal size was smaller as the cooling rate is faster. Although it was expectable that the decontamination performance was improved by diminishing the specific surface of the crystals, it was suggested that a large crystal produced by crystallization was not always high purity. Concerning the behavior of FPs, Eu behaved similarly to Pu(IV). Cs accompanied with U into the crystals under the condition in this test.     

Current status of researches on the plutonium rock-like oxide fuel and its burning in light water reactors

Intention of the ROX-LWR system research is to provide an option for utilization or disposition of surplus plutonium. Researches on inert matrix materials and irradiation performance shows that the most favorable candidate for the ROX fuel is a particle dispersed fuel where small particles consisted of yttria stabilized zirconia, PuO₂ and some additives are homogeneously dispersed in spinel matrix. Reactor safety analyses show that the ROX fueled PWR core has nearly the same performability as the existing UO₂ fueled PWR under both reactivity initiated accidents and loss of coolant accidents.     

VVER机组一回路硫酸根升高原因分析

针对核电厂大修期间一回路硫酸根异常升高的问题,首先对一回路和乏燃料水池可能产生硫酸根的物项成分进行分析,排除了给水、硼酸、氢氧化钾等添加试剂是造成硫酸根的主要来源。在对大修期间一回路硫酸根的变化趋势分析时发现,一回路硫酸根的变化和净化系统有关。通过试验确认乏燃料水池中的硼酸溶液在放射性和富氧条件下生成了氧化物质,当乏燃料水池和硼箱净化系统在净化乏燃料水池时,阳树脂中的磺酸基被氧化脱落进而分解生成硫酸根是导致一回路硫酸根升高的主要原因。根据研究成果通过减少阳床的运行时间有效解决了VVER机组中普遍存在的问题。     

Effect of re-oxidation rate of additive cations on corrosion rate of stainless steel in boiling nitric acid solution

The boiling nitric acid solution containing highly oxidizing cations dissolved from the spent nuclear fuels corrodes stainless steels because of the nobler corrosion potential and their fast reduction rate. The cations themselves are re-oxidized to higher oxidizing states in a bulk solution after the corrosion reaction. In this paper, the re-oxidation rate constants of typical cations, such as Cr, V, Pu, and Np, were analyzed, and discussed about the effect on time dependencies of the corrosion rate. It was indicated that the cations with a large re-oxidation rate constant, such as Np, could keep the corrosion rate at high level continuously for the long immersion duration.     

Separation of Antimony-125 in Fission Products

Most of the known radioactive nuclides of antimony produced by neutron irradiation of uranium have fission yields below 1% and have half-lives below 60 days. An exception is ¹²⁵Sb with a half-life of 2.7 yr, which raise its relative importance among the fission products with lapse of time after irradiation, and after 1 yr of cooling, its radioactivity is no longer negligible. This circumstance has led to its being separated from such sources as fall-out. No studies have so far been reported on using the nitrate system for this separation, though it is utilized in the reprocessing of spent fuel and in the dissolution of uranium samples. The present work describes a method of separating ¹²⁵Sb from fission products with use made of silica gel—nitric acid system, and an example of its application to the separation of ¹²⁵Sb from the spent fuel of JPDR-1. The fuel was irradiated from Oct. 1963 to Sep. 1969. The amount of ¹²⁵Sb measured after separation was (1.7± O.19)×10^?1Ci/gU at June 1972.     

Behavior of Ruthenium in Fluoride-Volatility Processes, (IV)

The products of the following three reactions were studied in relation to the reprocessing of oxide fuels: (i) fluorination of Ru by F₂ (ii) fluorination of Ru by a O₂-F₂ mixture, and (iii) secondary process of RuO₂-F₂ reaction. The product of Ru-F₂ reaction was only RuF₅; the mass spectrum of RuF₅ was obtained. Fluorination of Ru by a O₂-F₂ mixture resulted in the production of RuF₅ (85～75%) and RuOF₄ (15～25%); these results are different from those reported by earlier workers. The use of radioactive Ru^*O₂ traced the behavior of RuOF₄ in the apparatus. RuOF₄ decomposes on the wall which was not preliminarily coated with ruthenium; refluorination was effective for removal of the deposit. These results suggest that the fluorination of irradiated oxide fuels volatilizes the ruthenium as a mixture of RuOF₄, RuF₅, and a small amount of RuO₄.     

Sequential Ion-Exchange Separation of Heavy Elements and Selected Fission Products for Burnup Measurement

A sequential ion-exchange separation method was developed for use in burnup measurements of nuclear fuels. Group separation by anion-exchange resin column with hydrochloric acid solutions containing small amounts of nitric acid and hydrochloric acid was followed by various cation and anion- exchange processes. The heavy elements, such as U, Np and Pu, and some fission products selected as burnup monitors, such as Cs, Mo and Nd, could be sequentially and quantitatively separated from a sample taken from spent fuel. The recovery of these elements through the separation processes were examined. The sampling ratio of an aliquot in reference to the whole fuel specimen was determined by adding as sampling monitor a known amount of Cu to the sample during dissolution. The validity of the ion-exchange separation technique for routine analysis for burnup measurements is also discussed.     

Three Phase Region in Ternary System UO2-PUO2-PU2O3

The thermal decomposition of PuF₆ in the temperature range from 100° to 160°C was studied. At 100° and 120°C, no decomposition was detected. At 140° and 160°C, the rate of decomposition during reaction lasting up to 20 hr, and the dependence of the decomposition on the initial PuF₆ pressure were determined. The results are explained by assuming that the rate of decomposition is proportional to the 0.4-th power of PuF₆ pressure. The activation energy of the reaction was found from calculation to be 14.8 kcal/mole.