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).     

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.     

Sorption of U(VI) on the 4-Mercaptopyridine Self-Assembled Monolayer

The sorption of U(VI) on the 4-mercaptopyridine self-assembled monolayer (4-PyS-SAM) on Au(111) was studied by cyclic voltammetry. Cyclic voltammograms (CVs) of the 4-PyS-SAM working electrode were obtained by contact with 1mM UO₂(NO₃)₂ solution, 1mM UO₂(NO₃)₂ and 50mM acetic acid solution, or 1mM UO₂(NO₃)₂ and 50mM oxalic acid solution for 6 h at pH 4. The reduction current of U(VI) to U(V) was detected in the CV. The CV of the U(VI) associated 4-PyS-SAM after transport to U(VI)-free 0.1M NaClO₄ solution showed that the reduction current was detected in the cases of UO₂(NO₃)₂ and U(VI)-acetate, but not in the case of U(VI)-oxalate solution, indicating that U(VI) was adsorbed on the 4-PyS-SAM from the UO₂(NO₃)₂ and U(VI)-acetate solutions, but not from U(VI)-oxalate solution. These results suggest that stability of U(VI)-4-PyS-SAM is not so high that U(VI)-4-PyS-SAM cannot be formed in the presence of 50mM oxalate.     

Precipitation ability to U(IV) and stability of 1,3-dimethyl-2-imidazolidone for selective precipitation of U(VI) in nitric acid media

As a part of the investigation of precipitants with selectivity to U(VI) in nitric acid media, a preliminary study on the precipitation ability of 1,3-dimethyl-2-imidazolidone (DMI) to U(IV), a simulant of Pu(IV), was performed. DMI is a ring compound like N-n-butyl-2-pyrrolidone (NBP) which is one of the pyrrolidone derivatives (NRPs) and a promising precipitant for U(VI). While DMI is known to precipitate U(VI) from 3 mol dm⁻³ (=M) HNO₃, no precipitate was observed in the solution containing 0.15 M U(IV) and 3 M HNO₃ by adding DMI at the ratio of [DMI]/[U(IV)] = 5. This indicates that the selectivity of DMI to U(VI) than U(IV) is much higher compared with that of NBP.On the other hand, the stability of DMI under γ-ray irradiation and heating in HNO₃ solutions (≤4 M) was also examined to evaluate the applicability of DMI to the practical process, because gradual acid hydrolysis of DMI is inevitable due to the nature of the chemical structure. As a result, it was found that the stability is strongly affected by the concentration of HNO₃. Namely, very few DMI in 2 M HNO₃ underwent the ring-opening by the irradiation up to 220 kGy and heating at 50 °C up to 5 h, respectively, indicating that these treated samples may still hold the precipitation ability to U(VI). On the contrary, the cleavage of the ring of DMI in 4 M HNO₃ was found to proceed easily. From the above results, it was concluded that DMI may be a candidate as a selective precipitant for U(VI) in HNO₃ solutions up to ca. 2 M.     

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.     

Measurement of Average Cross Section for 233U(n, 2n)232U Reaction

Isotopically pure ²³³U samples, with only 3 × l0^?3 ppm²³²U content, were prepared by thermal neutron irradiation of thoria and subsequent chemical processing. The ²³³U sample thus obtained was reirradiated with a fission neutron spectrum in the core of the Kyoto University Reactor (KUR), and measurements were made of the fission spectrum average cross section for the ²³³U(n, 2n) ²³²U reaction. A value of 4.08±0.30 mb was obtained for this cross section, in agreement with the renormalized value of Halperin et al. within the limits of experimental error.

In order to assess the energy dependent cross section from the value of this integral measurement, the ²³³U (n, 2n) cross section was calculated assuming a Maxwellian-type fission spectrum and adopting the energy dependent evaluated cross sections of ENDF/B-III and other authors. The values of the cross section thus determined were found to be about 32 to 91% larger than the measured cross section given above. The result of Pearlstein's calculation of the ²³³U(n, 2n) cross section by the statistical model, again assuming the Maxwellian distribution, is smaller than the measured cross section by about 19%.     

Influence of solvents on extraction of U（VI） by N,N‘—didecanoylpiperazine

Extraction behavior of N,N‘-didecanoylpiperazine(DDPEZ)for U(VI) in a series of solvents from aqueous nitric acid media was investigated for the first time.The dependence of distribution ratios on the concentration of aqueous nitric acid,extractant and temperature has been discussed.The increasing sequence of extractive ability of DDPEZ is given:chloroform,carbon tetrachloride,dimethylbenzene,toluene,benzene.     

Development of a Multi-functional Reprocessing Process based on Ion-exchange Method by using Tertiary Pyridine-type Resin

A series of separation experiment was performed in order to study a multi-functional spent fuel reprocessing process based on ion-exchange technique. The tertiary pyridine-type anion-exchange resin was used in this experiment and the mixed oxide fuel highly irradiated in the experimental fast reactor “JOYO” was used as a reference spent fuel. As the result, ¹⁰⁶Ru+¹²⁵Sb, ¹³⁷Cs+¹⁵⁵Eu+¹⁴⁴Ce, plutonium, americium and curium could be separated from the irradiated fuel by only three steps of ion-exchange. The decontamination factor of ¹³⁷Cs and trivalent lanthanides (¹⁵⁵Eu, ¹⁴⁴Ce) in the final americium product exceeded 3.9x10⁴ and 1.0x10⁵, respectively. The decontamination factor for the mutual separation of ²⁴³Cm and ²⁴¹Am was larger than 2.2x10³ for the americium product and, moreover, the content of ¹³⁷Cs, trivalent lanthanides and ²⁴³Cm included in ²⁴¹Am product did not exceed 2 ppm.

These results prove that the proposed simplified separation process has a reality as a candidate for future reprocessing process based on the partitioning and transmutation concept.     

Influence of Gamma-Irradiation on Extraction and Cation-Transport of Uranium(VI) and Plutonium(IV) by Dicyclohexano-18-Crown-6

Changes of the extraction behaviour and cation-transport of U (VI) and Pu (IV) nitrates with r-irradiated DC18C6 in toluene have been investigated. The effect of radiation damage to DC18C6 for extraction and permeation at excessively higher doses (70 Mrads) has been studied systematically. No deterioration in its performance is noted at lower doses. Hydrolytic stability of the macrocycle/diluent system in the presence of nitric acid and radiolytic stability of immobilized liquid membrane using ‘Enka’ Accurel flat sheet polypropylene membrane films as solid support are also examined.     

A Method of Calculation on a Step Cascade with Side Flow for Enriching Uranium

The step cascade with side flow is defined as a step cascade where uranium of lower assays than that from the final stage is withdrawn, and where also, in addition to the major feed, reprocessed spent uranium is introduced as minor feed. A method of calculation is proposed for finding the exact total flow rate of the stream entering the separators in a step cascade with side flow in steady state and without losses of uranium.

Using this method of calculation, it is found that the separation factor of diffusers in the new gaseous diffusion plant to be established in the United States is about 1.0022. Moreover, the characteristics are revealed for a step cascade with side flow adopted for a plant treating the estimated amounts of enriched uranium necessary for and uranium discharged from pressurized and boiling water reactors in the nuclear power generating program of Japan in 1985. It has been disclosed from these results of calculation that the method presented here proposes useful means for a detailed design of the plant.     

A Nuclear Magnetic Resonance Study on Ligand Exchange Reactions in La(III), Nd(III), and U(VI) Nitrato Complexes with n-Octyl(pheny1)-N,N-Diisobutylcarbamoylmethylphosphine Oxide in Non-aqueous Solvents

The exchange reactions of n-octyl(pheny1)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in La(III), Nd(III), and U(VI) nitrate complexes with CMPO (La(III)-, Nd(III)-, and U(VI)-CMPO complexes) have been studied in CD₃COCD₃ by means of ³¹P NMR method. The number of CMPO coordinated to the first coordination sphere of La(III) ion was directly determined to be 3 by the area integrations of ³¹P NMR signals of free and coordinated CMPO molecules. The same coordination number of 3 was also obtained for the U(VI)-CMPO complex. The coordination number was not determined for the Nd(III)-CMPO complex, because of its paramagnetic behavior. The exchange rate constants of CMPO in La(III)- and U(VI)- CMPO complexes were obtained by the two-site exchange model. Paramagnetic line broadening was observed in the Nd(III)-CMPO complex and the rate constant for the exchange of CMPO was determined by the line-broadening method. The exchange rates of CMPO in La(III)- and Nd(III)-CMPO complexes depend on the free CMPO concentration ([CMPO]), while that in U(VI)-CMPO complex is independent of [CMPO]. The dissociative (D) and dissociative interchange (I_d ) mechanisms were proposed for the exchange reactions in the La(III)- and Nd(III)-CMPO complexes, and dissociative (D) or I_d mechanism was proposed for the U(VI)-CMPO complex. The dissociative rate constants (s^?1) at 25°C and activation parameters ΔH^# (kJ·mol^?1) and ΔS^# (J·K^?1·mol^?1) are 4.76x10³, 28.7±0.1, ?78.4±0.2 for La(III)-CMPO complex, 4.72x10³, 42.6±0.4, ?31.7±1.3 for Nd(III)-CMPO complex, and 3.20x10³, 46.9±0.6, ?20.5±2.2 for U(VI)-CMPO complex, respectively.     

Plasma-induced grafting of acrylic acid on bentonite for the removal of U(VI) from aqueous solution

Fabrication of reusable adsorbents with satisfactory adsorption capacity and using environmentfriendly preparation processes is required for the environment-related applications. In this study,acrylic acid(AA) was grafted onto bentonite(BT) to generate an AA-graft-BT(AA-g-BT)composite using a plasma-induced grafting technique considered to be an environment-friendly method. The as-prepared composite was characterized by scanning electron microscopy, x-ray powder diffraction, thermal gravity analysis, Fourier transform infrared spectroscopy and Barrett–Emmett–Teller analysis, demonstrating the successful grafting of AA onto BT. In addition, the removal of uranium(VI)(U(VI)) from contaminated aqueous solutions was examined using the as-prepared composite. The influencing factors, including contact time,p H value, ionic strength, temperature, and initial concentration, for the removal of U(VI) were investigated by batch experiments. The experimental process fitted best with the pseudo-secondorder kinetic and the Langmuir models. Moreover, thermodynamic investigation revealed a spontaneous and endothermic process. Compared with previous adsorbents, AA-g-BT has potential practical applications in treating U(VI)-contaminated solutions.     

Development of a suppression method for deposition of radioactive cobalt after chemical decontamination: (III) the suppression mechanism with preoxidized ferrite film for deposition of radioactive cobalt

The Hitachi ferrite coating film process (Hi-F) has been developed to lower recontamination after chemical decontamination. In this process, the chemical decontamination process is carried out, and a fine Fe₃O₄ coating film is formed on the surface of stainless steel piping in an aqueous solution. In order to improve the suppression of ⁶⁰Co deposition further, we combined the original Hi-F with a preoxidation step. We found the deposited amount of ⁶⁰Co with preoxidized Hi-F coating film (OHi-FC) was 1/10 of that for non-coated specimens. In this study, we investigated the suppression mechanism of ⁶⁰Co for the OHi-FC. The composition of OHi-FC was changed from Fe₃O₄ to Fe₂O₃ and then the crystals in the OHi-FC grew three times larger than those of the original Hi-F coating film. Consequently the corrosion amount of the stainless steel base metal was reduced by getting larger grains in the coating film. Because ⁶⁰Co was incorporated into the corrosion oxide, the suppression effect of ⁶⁰Co deposition by preoxidation was attributed to the suppression of the formation of the corrosion oxide by the OHi-FC.     

Development of a method to lower recontamination after chemical decontamination by depositing Pt nano particles: (II) consideration of the Pt effect on oxide composition

The Pt coating (Pt-C) process has been developed to lower the recontamination by ⁶⁰Co which was incorporated in oxides on piping surface after chemical decontamination. In order to determine the suppression mechanism of ⁶⁰Co deposition by Pt-C, it is important to investigate the formation of oxide film ⁶⁰Co deposition behavior on oxide with Pt-C specimens. In this paper, we observed the composition change of oxide after a ⁶⁰Co deposition test under the hydrogen water chemistry condition, and considered the ⁶⁰Co deposition behavior on oxide for Pt-C specimens. The Ni and Co metal concentrations in oxide were dramatically changed by Pt-C process. The Ni metal concentrations in oxide for specimens with and without the Pt-C process were 11.2% and 18.0%, respectively. On the other hand, the Co metal concentrations in oxide for specimens with and without the Pt-C process were 1.2% and 0.2%, respectively. This composition change of the oxides indicated that ⁶⁰Co incorporation for the Pt-C specimens was suppressed by replacing ⁶⁰Co with Ni. We concluded that the Ni²⁺ ions were incorporated into the 8a site of the oxide spinel structure instead of Co²⁺ ions due to the effect of the conversion deposition energy.     

用放射性示踪法研究分离铀和铅、铋

用放射性示踪法研究了TBP－苯体系对铀和铅、铋的萃取分离。实验结果表明,在6mol/LHNO3介质中用TBP－苯作为萃取剂可使铀和铅、铋达到定量分离。阴离子交换法对铀进一步纯化后,利用分子电镀法制得厚度为500μg/cm^2的^233U同位素靶。     

Application of N,N-di(2-ethylhexyl)butanamide for mutual separation of U(VI) and Pu(IV) by continuous counter-current extraction with mixer-settler extractors

Continuous counter-current extraction using N,N-di(2-ethylhexyl)butanamide (DEHBA) as an extractant was performed with mixer-settler type extractors consisting of U–Pu extraction, scrub, U recovery, Pu back-extraction, and U back-extraction steps. The feed solution used in the continuous counter-current extraction was 3 mol/dm³ (M) nitric acid containing U, Pu, and simulated fission products of Sr, Ba, Zr, Mo, Ru, Rh, Pd, and Nd. More than 99.9% of U and Pu in the feed was extracted by 1.9 M DEHBA at the U–Pu extraction step with negligible extraction of Sr, Ba, Mo, Ru, Rh, and Nd. The extracted Pu was back-extracted via contact with 0.3 M nitric acid in the Pu back-extraction step, and the ratio of Pu distributed to the Pu fraction stream was ～ 82%. It was confirmed that 1.9 M DEHBA effectively recovered U in the U recovery step, and the ratio of U in the Pu fraction stream was less than 1%. The extracted U was back-extracted in the U back-extraction step, and more than 98% of U was recovered in the U fraction stream.     

备用系统可靠性动态问题方法探究

随着核电安全性日益受到世界公众关注,用概率论的方法对核电厂系统进行可靠性分析也越来越显示出其重要作用.在进行系统可靠性分析时,故障树方法是国际上公认的一种简单、有效、经典的方法.但随着所要分析的工程系统日趋庞大,系统工作过程日趋复杂,对含有动态过程的系统可靠性分析成为故障树方法中的棘手问题.本文以可靠性分析中的典型动态问题--备用系统的可靠性为研究对象,运用蒙特卡罗方法编程探究如何运用故障树来求解可靠性动态问题,并在进行程序验证后对一个备用系统实例进行计算,给出计算结果.     

Cesium (Cs) particle formation based on a laser photochemical reaction with a self-injection-seeded Ti:sapphire laser for Cs isotope separation

We studied laser isotope separation of cesium (Cs) on the basis of a laser photochemical reaction and two-photon excitation scheme using a narrow line width Ti:sapphire laser to reduce long-term radioactive toxicity of a long-lived fission product. Using resonant laser irradiation to Cs atoms in hydrogen gas, we observed cesium hydride fine particles and confirmed the formation by calculations using rate equations. Our results show that the process seems promising for efficient Cs isotope separation.     

X(γ)刀吸收剂量测量方法研究

利用自制的直径f160mm有机玻璃球形模体和能够插入球形模体中心的PTW31006型0.015cm3尖点电离室测量了3种X(g)刀的等中心处不同辐射野的吸收剂量,并与半导体探测器(灵敏体积为1mm2?.5mm)的测量数据进行了比对,对电离室和半导体探测器的不同适用范围进行了分析,给出了一种适合X(g)刀的吸收剂量测量方法。