A study on selective precipitation ability of cyclic urea to U(VI) for developing reprocessing system based on precipitation method

We have proposed a new reprocessing system based on precipitation method. In order to find out precipitants with high selectivity to U(VI) and to investigate factors controlling precipitation ability to U(VI) and U(IV), properties of 3,4,5,6-tetrahydro-1,3-dimethyl-2(1H)-pyrimidinone (DMPU) as a precipitant have been examined by using U(VI), U(IV) as a simulant of Pu(IV), and simulated fission products (FPs). We have evaluated precipitation ratios (P.R.) for U(VI) and U(IV), solubility of U(VI) precipitates to 3.0 mol dm^?3 (M) HNO₃ solution, melting points (MPs) of U(VI) precipitates, log P (distribution ratio of a substance in 1-octanol/water biphasic system, a measure of hydrophobicity) of precipitants, and decontamination factors (DFs) of FPs. The properties of DMPU were compared with those in systems using N-n-butyl-2-pyrrolidone (NBP), N-cyclohexyl-2-pyrrolidone (NCP), and other pyrrolidone derivatives as the precipitant. The P.R. values of DMPU to U(VI) and U(IV) in 3.0 M HNO₃ solutions were around 99% at [DMPU]/[U(VI)] = 2.0 and 0% at [DMPU]/[U(IV)] = 5.0, respectively. In DMPU system, the DF values of the most of simulated FPs [Rb(I), Cs(I), Sr(II), Ba(II), Ru(III), Rh(III), La(III), Ce(III), Pr(III), Nd(III), and Sm(III)] used in the present study were found to be more than 100. Even in U(VI)–U(IV) coexisting system, the selectivity of DMPU to U(VI) was higher than those of NBP and NCP. This selective precipitation ability of DMPU to U(VI) was evaluated by the solubility of U(VI) precipitates on the basis of their MPs and the log P values of precipitants. As a result, it was found that the precipitants having low hydrophobicity and forming the U(VI) precipitates with high MPs have highly selective precipitation ability to U(VI).     

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.     

Mutual Separation of U,Pu and Fission Products with Tri-iso-octylamine

Uranium and Pu were extracted with tri-wo-octylamine (CC1₄) from irradiated U dissolved in sulfuric acid, from which Pu was back-extracted with sulfurous acid, and U with I M NH₄AC-1.5%H₂O₂. The decontamination factors were estimated to be 10² for U fraction (γ and β) activities) and >10³ for Pu fraction.     

低浓缩铀靶辐照后裂变核素产量的理论研究

用低浓缩铀靶代替高浓缩铀靶辐照进行⁹⁹Mo、¹³¹I等医用放射性核素生产是一个必然的趋势。本文利用输运计算程序DRAGON研究了靶件²³⁵U富集度、中子注量率、辐照时间对⁹⁹Mo、¹³¹I、⁹⁰Sr、⁹⁵Zr、²³⁹Pu等核素比活度变化的影响,以及不同²³⁵U富集度下裂变体系组成和总比活度的变化规律。计算结果表明,本文考察的10余种核素比活度的变化随辐照时间的不同而有所不同,其中⁹⁹Mo、¹³¹I、¹⁴⁷Nd和¹³³Xe等核素的比活度可快速达到饱和,⁸⁹Sr、¹⁰³Ru、⁹⁵Zr和¹⁴¹Ce等缓慢达到饱和,而⁹⁹Tc、⁸⁵Kr和⁹⁰Sr、²³⁹Pu在计算时间内达不到饱和,但所有核素的比活度随时间的变化趋势与靶件²³⁵U富集度无关;⁹⁹Mo、¹³¹I、⁹⁰Sr、⁹⁵Zr等核素的比活度均随靶件²³⁵U富集度提高而增加,而²³⁹Pu比活度则随着靶件富集度的减少而显著增加,提示改用低浓缩铀靶进行⁹⁹Mo、¹³¹I等医用放射性核素生产时应特别关注²³⁹Pu带来的影响;核素比活度随中子注量率的增加而线性增加,且斜率基本相同;靶件辐照时间的改变不会明显影响裂变体系的组成,在低浓缩铀（²³⁵U含量≤20%）区域,靶件²³⁵U富集度对裂变体系的组成影响很小。     

Liquid Metal Extraction for Removal of Molybdenum from Molten Glass Containing Simulated Nuclear Waste Elements

Laboratory-scale experiments for removing Mo and MoO3 from molten borosilicate glass were performed using liquid Cu as an extractant. Removal of Mo from the simulated HLW glass containing oxides of Nd, Fe, Zr, Mo, Sn, Ni, Sr, Cd, Ru, and Se was also performed, and the fractions of these elements transferred into Cu were examined. Mixtures of Cu anda ternary SiO₂-B₂O₃-Na₂O glass containing metallic Mo or MoO₃ were heated in an alumina crucible at 1,673K in an Ar environment. The amounts of Mo and MoO₃ added to 10 g of the ternary glass were fixed at 0.1 and 0.15 g, respectively. As for the glass containing metallic Mo, more than 90% of Mo was extracted into liquid Cu. Spherical Cu metal buttons containing Mo formed on the bottom of the crucible when Cu was added at more than 10 times that of Mo on a mass basis. Removal of Mo from the glass containing MoO3 was also achieved by the addition of Si as a reducing agent for the reduction from MoO₃ to Mo. The fraction of Mo extracted into liquid Cu depended on the molar ratio of Si to Cu added to the glass. The fraction increased up to 84% with an increase in the molar ratio of Si/Cu. However, the excess addition of Si may enhance the chemical interaction between the metal phase and the glass phase, and some of the metal phase containing Mo remained in the glass phase without forming a metal button. The optimum molar ratio of Si/Cu that produces the highest removal fraction was found to be approximately 0.5. Almost the same removal fraction of 88% was obtained from the simulated HLW glass under the condition of Si/Cu = 0.5. Nearly 100% of Ru was extracted into Cu with Mo, while Sr, Zr, and Nd were hardly extracted and remained in the glass.     

Composition and structure of fission product precipitates in irradiated oxide fuels: Correlation with phase studies in the Mo-Ru-Rh-Pd and BaO-UO2-ZrO2-MoO2 Systems

Composition and crystal structure of fission product precipitates in irradiated oxide fuels were studied by X-ray microanalysis and X-ray diffraction using instruments shielded for α-contamination and β-γ-radiation. Pin cross-sections, fuel micro-samples from 300 μm hollow drillings and residues from the dissolution of irradiated material in HNO₃ were investigated. The metallic phases found are hcp ?-Ru(Mo,Tc,Rh,Pd) solid solutions with broad variations in concentration of the components, bcc β-Mo(Tc,Ru) and fcc α-Pd(Ru,Rh). The dominating ceramic precipitate is composed of ($\text{Ba}{}_{\mathrm{1-x-y}}\text{Sr}{}_{\mathrm{x}}\text{Cs}{}_{\mathrm{y}}$)(U,Pu,RE,Zr,Mo)O₃ which crystallizes in the cubic perovskite type. The Mo fraction of these phases is related to the local oxygen potential of the fuels. The in-pile observed results agree well with phase studies in the quaternary Mo-Ru-Rh-Pd system where complete solid solubility exists between hcp Ru and the hexagonally stabilized Mo-Rh and Mo-Pd phases. Agreement is further attained with phase studies in the pseudoquaternary BaO-UO₂-ZrO₂-MoO₂ system which is characterized by a cubic perovskite phase Ba(U,Zr,Mo)O₃.     

Low-Temperature Partitioning of Plutonium with High U(VI)-Loading for Fuel Reprocessing

Measurement of the distribution ratios of Pu(IV), U(VI) and HNO₃ at low temperatures and its treatment with DIST code revealed that a high U (VI)-loading of 30% TBP in n-dodecane splits Pu(IV) down to the aqueous phase more strongly than do at 25°C. Based on these findings, flowsheet conditions to separate Pu(IV) from U(VI) were investigated with EXTRA.M code including the distribution equations obtained above. And tentative flowsheets for non-reductive Pu-splitting process at a temperature of 5°C were proposed for fuel reprocessing mainly based on the effects of U (VI)-loading in the solvent and temperature on distribution ratios of Pu(IV) and U(VI). Distribution ratios of the fission products, Zr, Nb, Ru and Ce were also measured to assess their decontamination from U or Pu products in the above process. Finally behavior of Np, in the proposed partitioning process was discussed by analysis with EXTRA. M code and a redox reaction model.     

Studies on Focusing Chromatography

U, Np and F.P. in neutron irradiated uranium oxide were separated from each other using focusing chromatography. Individual nuclides were determined hy γ-ray spectrometry. As complex forming agent, 1.0M lactic acid and 0.05M nitrilotriacetic acid (NTA) were tried, and various mixtures of hydrochloric acid and sodium chloride as complex decomposing agent.

The efficacy of the forming and decomposing agents for the separation was observed with various solutions of electrolytes. In all cases, Tc, Mo, Ru, Rh, I, Zr and one fraction of Np migrated toward the anode. As the hydrogen ion concentration of the complex decomposing solution was decreased, these nuclides became more easily separable. Te and a small amount of Np generally remained in the area of the spot where the sample had been introduced. Sr and Ba always migrated toward the cathode. Rare earth elements were sensitive not only to the kind of complex forming agent used but also to the pH value of the decomposing solution. Np manifested three different modes of focusing characteristic, their proportions varying with the conditions of the separation system.

The concluding result was that the system of 0.1M sodium chloride—1.0 M lactic acid was found the most suitable for separating these multi-component mixtures.     

Equilibrium and kinetic studies of selective adsorption and separation for strontium using DtBuCH18C6 loaded resin

A crown ether loaded resin was prepared by successive impregnation and fixing the 4′,4′(5″)-di(tert-butylcyclohexano)-18-crown-6 (DtBuCH18C6) and its molecule modifier, 1-dodecanol, onto the porous silica/polymer composite support (SiO₂-P particles). The characterization of DtBuCH18C6 loaded resin was examined by thermal gravimetry and differential thermal analysis and electron probe microanalysis. The adsorption behavior of Sr(II), Cs(I), Ru(III), Pd(II), La(III), Nd(III), Sm(III), Gd(III), Zr(IV), and Mo(VI) was investigated by the batch method. Furthermore, the column test for Sr (II) was performed. The batch experiments were carried out by varying the shaking times, HNO₃ concentration, and initial concentration of metal ions. A relatively large K _d value above 182 cm³/g for Sr(II) was obtained in the presence of 3 M HNO₃. In contrast, the K _d values of Cs(I), Ru(III), Pd(II), La(III), Nd(III), Sm(III), Gd(III), Zr(IV), and Mo(VI) were considerably lower than 10 cm³/g. The adsorption of Sr(II) was found to be controlled by chemisorption mechanism, and followed a Langmuir-type adsorption equation. The breakthrough curve of Sr(II) had S-shaped profile, and the elution percentage was estimated to be 99.9% by using the eluent of H₂O.     

Pressure dependence of extraction behavior of plutonium(IV) and uranium(VI) from nitric acid solution to supercritical carbon dioxide containing tributylphosphate

Plutonium(IV) and uranium(VI) were extracted into supercritical CO₂ fluid phase (SF-CO₂) containing tributylphosphate (TBP) with equilibrium distribution ratios, D, e. g., D_Pu(IV) = 3.1 and D_U(IV) = 2.0, for the extraction of 2 × 10⁻³ M Pu(IV) and U(VI) from 3 M HNO₃ into SF-CO₂ containing 0.3 M TBP at 60 °C and 15 MPa. A simple linear relation between D and density of CO₂; long D = a log + b (a,b; constants), was observed, which was explained theoretically by the formulation of the extraction equilibrium taking into account the phase distribution behavior of extractant TBP and extracted species, i.e. Pu(IV)- and U(VI)-TBP complexes involved in the extraction reaction. The slopes a of the log D vs. log plots were −(1.6 ± 0.1) and −(2.7 ± 0.5) for the extraction of Pu(IV) and U(VI), respectively. The differences in D as well as the slope a between Pu(IV) and U(IV) make it possible to design the U and Pu separation method by which one can achieve an enhancement of the extraction efficiency and selectivity by tuning the operation pressure.     

Lattice physics analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel

As part of a validation study of burnup calculations of BWR cores, lattice physics analyses were performed on burnups and isotopic compositions of U, Pu and fission product nuclides measured on five samples taken from 9 × 9 BWR fuel assemblies. Burnup calculations in infinite assembly geometry were carried out using MVP-BURN and SRAC codes coupled with major nuclear data libraries. The burnups determined based on the Nd-148 method were from 27.9 to 64.2 GWd/t. The typical relative differences in isotopic compositions (atom/Total-U) between the burnup calculations and measurements were ?2 ～ 19% for ²³⁴U, ?20 ～ 3% for ²³⁵U, ?1.5 ～ 0.1% for ²³⁶U, ?0.04 ～ 0.02% for ²³⁸U, ?4 ～ 11% for ²³⁸Pu, ?11 ～ ?2% for ²³⁹Pu, ?3 ～ 0% for ²⁴⁰Pu, ?12 ～ ?2% for ²⁴¹Pu and ?2 ～ 3% for ²⁴²Pu. They were ?2 ～ 2% for Nd isotopes, ?15 ～ 7% for Eu isotopes, ?13 ～ 1% for Cs isotopes, ?13 ～ 8% for Sm isotopes, 0 ～ 7% for ¹⁴⁷Pm, ?7 ～ ?2% for ⁹⁵Mo, ?2 ～ ?1% for¹⁰¹Ru and 0 ～ 4% for ¹⁰³Rh.     

Radioactivity of fission product and heavy nuclides deposited on soil in Fukushima Dai-Ichi Nuclear Power Plant accident

Based on periodically performed radioactivity measurements on soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, activity ratios to ¹³⁷Cs of fission product and heavy nuclides were obtained for Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Ba, La, Pu, Am, and Cm isotopes. By exponentially fitting or averaging, the activity ratios at the core shutdown were estimated. Using correlations of activity ratios of ¹³⁴Cs to ¹³⁷Cs, and ²³⁸Pu to the sum of ²³⁹Pu and ²⁴⁰Pu against fuel burnup, burnup of the fuel sourcing the deposited activity of the soil was estimated. The activity ratios to ¹³⁷Cs of each nuclide on the deposited activity were divided by those calculated on the fuel at the shutdown to obtain the deposited activity fraction of each nuclide as a relative value to ¹³⁷Cs, which also corresponds to the deposited fraction of each element as a relative value to Cs. The obtained deposited fractions relative to Cs are the orders of 10^?4 to 10^?2 for Sr, 10^?5 to 10^?3 for Nb, 10^?2 to 10^?1 for Mo, 1 to 10 for I, 10^?3 to 10^?2 for Ba, 10^?2 for La, 10^?6 to 10^?3 for Pu, 10^?6 to 10^?4 for Am, and 10^?7 to 10^?5 for Cm. The deposited fractions for Tc, Ag, and Te were not estimated due to the lack of the calculated inventories in the fuel for the relevant measured radioactive nuclides.     

Release of radioactive materials from high active liquid waste in small-scale hot test for boiling accident in reprocessing plant

The release behavior of radioactive materials from high active liquid waste (HALW) has been experimentally investigated under boiling accident conditions. In the experiments using HALW obtained through laboratory-scale reprocessing, the release ratio was measured for fission product (FP) nuclides such as Ru, Tc-99, Cs, Sr, Nd, Y, Mo, Rh and actinides such as Cm-242 and Am-241. As a result, the release ratio was 0.20 for Ru and was around 1×10^?4 for the FP and actinide nuclides. Ru was released into the gas phase in the form of both mist and gas. For its released amount, weak dependency was found to its initial concentration in the test solution. The release ratio decreased with the increase in the initial concentration. For other FP nuclides and actinides as non-volatile, released into the gas phase in the form of mist, the released amount increased with the increase in the initial concentration. The release ratio of Ru and NOx concentration increased with the increase in the temperature of the test solutions. They were released together almost at the same temperature between 200 and 300 °C. Size distribution of particles like mist was measured. The data show that there was a difference between distributions at the temperatures below 150 °C and over 200 °C.     

Separation of Plutonium-238 from Fission Products by Solvent Extraction Using HDEHP

A procedure for separating ²³⁸Pu from a Np sample irradiated with neutrons is described. Rapid separation of Pu by HDEHP solvent extraction was attempted, and without adjusting its valency states in the dissolver solution of the sample. Both Pu(IV) and Pu(VI) were extracted along with Np from the HNO₃ solutions of various concentrations. The Pu and Np extracted in the organic solution were back-extracted with oxalic acid solutions. The decontamination factors of the crude products were of the order of 10² for gross γ-activity. The Pu in the products was separated from Np by means of ion exchange resin columns. Approximately 0.5 mg of ²³⁸Pu was obtained with an efficiency exceeding 95%.     

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO₃ concentrations. The flowsheet with a supply of 0.15 mol/dm³ HNO₃ solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×10⁵ and 2.42×10⁵, respectively.     

萃取液闪法测定生产堆高放废液及其处理工艺中的^99Tc

本文在ＨＮＯ３和Ｈ２ＳＯ２混合介质中，用环己酮从高放废液及其处理 样品中定量萃取＾９９Ｔｃ，分别以Ｋ２ＣＯ３－Ｈ２Ｏ２及ＮａＣＯ２洗涤有机相去除钌，铑，碘等核素。有机相与溶水性的闪烁液混匀，液闪法测量＾９９Ｔｃ的活度。     

效率示踪—液闪法测定后处理厂中^99Tc

研究了萃取－液闪技术快速测定９９Ｔｃ的方法。用环己酮作萃取剂，在Ｈ２ＳＯ４介质下从水相中选择性萃取Ｔｃ，使Ｔｃ与样品中的Ｕ、Ｐｕ、Ａｍ、Ｅｕ、Ｓｒ、Ｃｓ、Ｒｕ、Ｉ等干扰元素分离，９９Ｔｃ的萃取率大于９８％，对干扰元素的分离系数大于１０４。对３０％ＴＢＰ－煤油有机相样品，可用Ｈ２Ｏ反萃后转成Ｈ２ＳＯ４介质。萃取后的有机相直接加入闪烁液中测量，以效率示踪法确定测量效率，得到９９Ｔｃ活度。本方法可用于后处理工艺或废液样品中９９Ｔｃ的测定。     

用TRPO从高放废液中去除锕系元素的研究──微型离心萃取器串级冷实验

采用微型离心萃取器进行了ＴＲＰＯ流程从模拟高放废液中去除锕系元素的冷实验。实验中用Ｎｄ代替Ａｍ，Ｚｒ代替Ｎｐ、Ｐｕ，在模拟高放废液稀释３倍、酸度为１．０ｍｏｌ／ｌ时，采用１２级萃取、４级洗涤能有效地去除模拟高放废液中９９．９％以上的Ｎｄ、Ｚｒ，满足了冷实验要求，并且萃取中不出现三相，可以使萃入的Ｆｅ洗下６０％，避免大量Ｆｅ进入后续流程。采用硝酸、草酸分别反萃Ｎｄ和Ｚｒ，使Ｎｄ和Ｚｒ分成二组，交叉污染很小。文中给出了硝酸、Ｎｄ、Ｚｒ等在各级的浓度剖面和它们在各物流中的分布。     

Investigation of the partitioning of fission products between the metallic and oxide phases of melt in a VVER-1000 vessel during a serious accident

The results of investigations of the interaction of U-Zr-B-C-O melt with steel, performed as part of the OECD MASCA program, are presented. It is established that the presence of Mo, Ru, Sr, Ba, Ce, and La in the melt does not qualitatively affect the interaction with structural steel and the character of the stratification of the melt in the reactor vessel. The partition factors of the fission products between the oxide and metallic phases are determined as a function of the oxidation of the melt, the ratio U/Zr, the composition of the structural steel, and the temperature. __________ Translated from Atomnaya énergiya, Vol. 105, No. 1, pp. 3–7, July, 2008.     

Selective uptake properties of metal ions by hybrid microcapsules enclosed with TBP

The granulation of TBP extractant is effective for the enhancement of uptake efficiency. The granulation was accomplished by microencapsulating techniques using alginate gel polymers (alginate and alginic acid gel polymers; calcium alginate, barium alginate and nitric alginate (CaALG, BaALG and HALG)). The characterization of hybrid microcapsules was examined by SEM/EPMA, and the uptake properties and the selectivity of various nuclides, Fe(III), Sr(II), Co(II), U(VI) and Pu(IV), were examined by batch methods. A relatively high uptake (%) of Fe(III), Sr(II) and Co(II) above 80% was obtained in the presence of 10⁻³ M HNO₃, and the uptake equilibrium was attained within 5 h. The uptake rate of U(VI) and Pu(IV) attained equilibrium within 1 h and 3 h, respectively. At higher HNO₃ concentration ranging from 10⁻³ M to 5 M, the uptake (%) of Fe(III), Sr(II) and Co(II) was considerably lowered. In contrast, the uptake (%) of U(VI) and Pu(IV) about 60% was obtained even in the presence of 5 M HNO₃. The uptake of U(VI) for MCs (TBP–CaALG) was governed by the extraction with TBP micro droplets and ion-exchange reaction in the CaALG matrices. Energy dispersive spectra (EDS) showed that U(VI) ions were incorporated into both phases of TBP and CaALG in microcapsules.