Photochemically-Induced Valency Adjustment of Plutonium and Neptunium in Nitric Acid Solution Using Mercury Lamp

A photochemically-induced valency adjustment method has been studied to remove Np from the mixed nitric acid solutions of Pu and Np in connection with the Purex reprocessing. The valencies of Pu and Np ions were adjusted to be Pu(HI) and Np(V) under the initial conditions and their concentrations were 1x10^?4 and 1x10^?3 mol·dm^?3, respectively. The experiments were carried out under the various conditions changing the irradiation intensities of the Hg lamp in the various concentrations of HNO₃. It was found that the rates of the redox reactions of the Pu ions were significantly affected by the irradiated light as well as the acid strength. Under the irradiation of the 0.015 W Hg lamp in 3 M HNO₃ solution containing a tenfold excess of a hydroxylamine and hydrazine, more than 95% Pu(ID) was oxidized rapidly to Pu(IV) within 10 min irradiation and it remained at the same valency even after the continuous further irradiation.

On the other hand, the irradiation did not change the valency of Np(V) under the conditions studied. These valency conditions, i.e. Pu(IV) and Np(V), are appropriate for separating Np from Pu by the solvent extraction with TBP-n-dodecane.

The present results lead to the conclusion that the photochemical method has a high potential for removing Np from the mixed solution of Pu and Np. The photochemical redox reaction mechanisms of Pu and Np in the nitric acid solution were discussed from the stand-points of the thermodynamic and kinetic considerations related to the variation in their standard electrode potentials of the photo-excited ion species by the light irradiation.     

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%.     

Controlling Neptunium and Plutonium within Single Cycle Solvent Extraction Flowsheets for Advanced Fuel Cycles

Simple hydroxamic acids are shown to be useful reagents for the separation of Np and Pu from U within simplified, single cycle Purex flowsheets. They are compatible with the use of centrifugal contactors and laboratory scale flowsheet trials with aceto-hydroxamic acid have demonstrated high actinide recoveries and decontamination factors on products for active feeds of up to 40 wt.% Pu. They therefore show many ideal characteristics for Pu and Np recovery within flowsheet options for actinide recovery in advanced fuel cycles. Furthermore, in order to optimize the routing of Np with the Pu product in advanced flowsheets, additional studies of Np extraction in the primary co-decontamination contactor, prior to U/Pu partition, have been undertaken, combining experiment, modelling and flowsheet tests.     

Numerical Simulation for Chemical Reactions of Actinide Elements in Aqueous Nitric Acid Solution

A computer code REACT incorporating 30 rate equations of reactions, i.e. radiolytic formation and decomposition of HNO₂, redox and disproportionation reactions, was developed to simulate behavior of actinide elements in the aqueous nitric acid solution. Main aspects of REACT code were explained briefly and then calculated results were compared with reported data to evaluate the model in the systems of radiolytic accumulation of HNO₂, stabilization process of Pu solution. The study showed that some radiolytic products other than HNO₂ would play a significant role and should be taken into account for precise simulation of very slow valency change of Pu in the neat Pu solution particularly with high radiation power density. Some examples of calculation were also shown for systems of reduction of Pp and Np by uranous or HAN and oxidation of Np (V) to Np (VI).     

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.     

1AW溶液中U、Np和Pu的分析

研究了H2O2同时调节镎、钚、铀价态至Np(Ⅳ)、Pu(Ⅳ)以及U(Ⅵ)的条件,在6mol/L HNO3浓度下,使用1.5%H2O2作为氧化还原剂对1AW进行调价,吸附上UTEVA柱并淋洗后,对钚、镎和铀进行洗脱。对模拟放射性样品进行预处理后,其中铀、镎、钚单独顺序洗脱的回收率分别为91.5%、119%、99.8%,137 Cs的去污因子高达7.4×104,单个样品操作时间约为1~1.5h;若钚洗脱后铀、镎同时洗脱并使用ED-XRF测量可以减少操作时间,铀、镎的回收率分别为102.4%、93.9%。均满足样品分析及辐射防护要求。     

Research and Development of Minor Actinide-containing Fuel and Target in a Future Integrated Closed Cycle System

Research and development of minor actinide-containing fuels and targets, i.e., (Pu,Am)O₂–MgO, (Pu,Np)O₂–MgO, (U,Pu,Np)O₂, (U,Pu,Np)N and (Pu,Np,Zr)N, for use in a future integrated closed cycle system that includes fast reactor and accelerator driven sub-critical system is underway. The present statuses of fabrication test and property measurements are given. Design concept of the oxide target is described in detail together with a screening of the support material. A new apparatus for the measurement of mechanical properties at the elevated temperature is installed for use in evaluating the fuel-cladding mechanical interaction. Development histories with future prospects of two types of Np-containing fuels for the fast reactor are mentioned. Preliminary test results for a new nitride target for the accelerator driven sub-critical system are given. Finally, an irradiation test plan in the experimental fast reactor JOYO is briefly described.     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.     

Constituent redistribution in U-Pu-Zr fuel during irradiation

The thermo-migration fluxes of U, Pu and Zr in U-Pu-Zr metallic alloy fuel during irradiation in the Experimental Breeder Reactor II (EBR-II) were calculated using the constituent redistribution profiles measured in postirradiation examinations. Based on these fluxes, the diffusion coefficients, and the sums of heat of transport and enthalpy of solution for the γ, γ+ζ and δ+ζ phases in U-Pu-Zr were obtained. Using these data, the predicted concentration redistribution profiles are consistent with the measurements. The effect of minor actinide (Am and Np) addition was also examined. Am migration generally followed that of Zr with local precipitation, while Np behaved similarly to Pu.     

用于锕系元素提取分离的萃取剂——TiAP

对磷酸三异戊酯(TiAP)和磷酸三丁酯(TBP)萃取体系的物理性质、萃取能力、耐辐照等方面进行了比较,结果表明,TiAP作为萃取剂在物理性质、萃取Pu(Ⅳ)和Np(Ⅳ)的能力以及辐照稳定性能等方面明显好于TBP。提出TiAP是一种很好的并有可能用于锕系元素提取分离的萃取剂。     

Actinide valences in xenotime and monazite

Tetravalent U, Np and Pu can be substituted by ceramic methods into the rare earth site of xenotime and monazite in air atmospheres using Ca ions as charge compensators, while no evidence of penta- or hexavalent actinide ions was found. Some Pu³⁺ and Np³⁺ can be incorporated in xenotime samples fired in a reducing atmosphere.     

普雷克斯流程中Np的行为和控制

在后处理流程的众多化学分离中 ,Np的走向和控制是国际后处理界关注的重点研究课题。根据我国和其他国家的研究成果 ,综合分析了后处理中Np的走向和控制。Np在辐照燃料溶解液中的价态分布主要取决于溶解液中HNO3与HNO2 之比 ,通常情况下 ,溶解液中Np(Ⅴ )占主要份额 ;Np在共去污阶段的走向有两种可能 ,一是将Np控制为Np(Ⅴ ) ,使其进入高放废液 (1AW ) ,二是将Np控制为Np(Ⅵ ) ,则Np将与U ,Pu一起进入有机相 ,但两者至今为止都难以实现定量分离。Np在U/Pu分离阶段部分随U ,部分随Pu。在U纯化循环中 ,理想的方法是采用低酸加热氧化Np(Ⅳ )至Np(Ⅴ ) ,以实现与铀的有效分离。     

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.     

甲基肼还原反萃取分离镎/钚的研究

镎的提取和分离是后处理领域重点关注的研究课题之一。甲基肼作为一种有机无盐试剂,其还原Np(Ⅵ)的速率快于还原Pu(Ⅳ)的速率,理论上可以利用其反应速率上的差异来实现镎与钚的分离。为了探索甲基肼还原反萃分离镎、钚的可行性,本文采用单级萃取池研究了甲基肼还原反萃Np(Ⅵ)和Pu(Ⅳ)的过程。通过考察还原剂浓度、硝酸浓度以及反应温度和搅拌速率等条件对甲基肼还原反萃Np(Ⅵ)和Pu(Ⅳ)过程的影响,确定了Np(Ⅵ)和Pu(Ⅳ)反萃动力学方程和表观活化能。通过所得的动力学方程得出甲基肼还原反萃Np(Ⅵ)和Pu(Ⅳ)的半反应时间,并对Np(Ⅵ)和Pu(Ⅳ)分离过程的工艺进行了初步探索。     

Neptunium Valence Adjustment through Photochemically Induced Redox Reactions at Low Concentrations

Photochemically induced redox reactions of Np(VI) and Np (IV) to Np (V) are experimentally studied in nitric acid solution of low Np concentration less than 10^?4 mo].dm⁸ using a Kr-F excimer laser. The preparations of the initial solution of Np (VI) or Np (IV) are based on the addition of chemical redox reagents. The extraction chromatography is used to analyze the fraction of the neptunium valences. The result of redox experiments with laser radiation shows that the fractions of Np(V)/Np(VI) or Np(V)/Np(IV) photochemically increase up to the steady-state values against initially oxidizing or reducing conditions respectively. The steady- state values are different from those at thermal equilibrium states. It is concluded that, in low Np concentration as is observed in the normal Purex process, laser application to valence adjustment of Np gives another redox condition which is different from that determined thermochemically.     

Oxidizing dissolution of spent MOX47 fuel subjected to water radiolysis: Solution chemistry and surface characterization by Raman spectroscopy

The mechanisms of oxidizing dissolution of spent MOX fuel (MIMAS TU2®) subjected to water radiolysis were investigated experimentally by leaching spent MOX47 fuel samples in pure water at 25 °C under different oxidizing conditions (with and without external gamma irradiation); the leached surfaces were characterized by Raman spectroscopy. The highly oxidizing conditions resulting from external gamma irradiation significantly increased the concentration of plutonium (Pu(V)) and uranium (U(VI)) compared with a benchmark experiment (without external irradiation). The oxidation behavior of the plutonium-enriched aggregates differed significantly from that of the UO₂ matrix after several months of leaching in water under gamma irradiation. The plutonium in the aggregates appears to limit fuel oxidation. The only secondary phases formed and identified to date by Raman spectroscopy are uranium peroxides that generally precipitate on the surface of the UO₂ grains. Concerning the behavior of plutonium, solution analysis results appear to be compatible with a conventional explanation based on an equilibrium with a Pu(OH)_4(am) phase. The fission product release - considered as a general indicator of matrix alteration - from MOX47 fuel also increases under external gamma irradiation and a change in the leaching mode is observed. Diffusive leaching was clearly identified, coinciding with the rapid onset of steady-state actinide concentrations in the bulk solution.     

肼衍生物与Pu（IV）的氧化还原反应定量构效关系研究

钚是最重要的放射性核素之一,Pu(IV)与有机还原的的氧化还原反应一直是核燃料后处理研究的重点之一。通过肼衍生物结构与Pu(IV)还原速率之间构效关系的研究,获得其关键影响因素,为镎钚分离提供关键数据及研究方向指引。采用密度泛函B3LYP方法和6-311+(3d,3p)基组对11种肼衍生物进行了几何优化和能量计算,获得了其最稳定构型。并利用HyperChem软件包计算优化后的分子的疏水性参数等结构描述符。运用数学统计软件SPSS对相应的物理化学参数进行相关性分析及逐步回归分析,最终得到具有良好相关性的构效关系(Quantitative structure-activity relationships, QSAR)方程。方程表明,疏水性参数是影响肼衍生物对Pu(IV)还原速率的主要因素,且与Pu(IV)还原速率呈负相关。     

Environmentally acceptable nuclear fuel cycle development of a new reprocessing system

The aim of the present study is to establish a new reprocessing system for spent nuclear fuel, which would overcome the environmental problems in the nuclear fuel cycle. In order to achieve this, the following subjects have to be conquered: recoveries of high ratios of uranium and trans uranium elements from spent nuclear fuel, separations of strong radioactive elements, such as Sr and Cs, and assurance of the extreme safety during operation. The last subjects might be of particular importance in order to avoid any potential danger. Therefore, in the present system all processes were performed under mild aqueous conditions. Experiments were carried out for a simulated spent fuel solution, which was calculated from the ORIGEN CODE containing uranium and 17 major elements. The system consists of the following processes: 1. dissolution of spent UO₂ fuel involving off-gas treatment of I and Ru; 2. neutralization of the dissolved fuel solution with NaHCO₃---Na₂CO₃ mixed solution to be slightly basic at pH about 9 followed by the separation of precipitated fission products by centrifugation; 3. separation of Cs by a precipitation method using tetraphenylborate ion; 4. recovery of U, Np and Pu as precipitates of hydrolyzed compounds from alkaline solution; 5. separation of Am and Cm from lanthanide elements. The concentration of residual uranium in the final solution was measured to be ppm order, indicating that the decontamination factor of U was as large as 10⁴. Other hexa-valent actinide ions, NpO₂²⁺ and PuO₂²⁺, also have extremely large stability constants for the complex formation with carbonate ion, and are expected to behave similarly with UO₂²⁺. In conclusion, the present reprocessing system enables us to recover U, Pu and Np from spent nuclear fuel by means of a simple precipitation method in much higher ratios compared with other reprocessing methods, to separate hazardous Cs and Sr from high-level waste, and to exclude any potential danger owing to chemical processes under mild aqueous conditions.     

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.     

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.