Behavior of Iodine-131 in Dissolution of Irradiated Uranium Dioxide

A hundred milligram of UO₂ was irradiated In JRR-4 (burn up 0.017%) and dissolved In boiling 3 M-HNO₃ (20 ml) to study the behavior of fission-product iodine (¹³¹I, ? 1 ng) During the dissolution, 80–90% of the iodine volatilized together with ¹³³Xe. Zeolite 13X (–4g) trapped completely the iodine in the dissolver off-gas Blowing a NO₂ flow through the solution was effective for expulsion of the remaining iodine and finally 90–95% of the initial iodine was removed from the solution with the aid of addition of KIO₃. The Iodine species in the solution includes I₂, I^?, IO₃ ^? (IO₄ ^?), organic iodides and other unknown species The expulsion operations caused the unknown species to increase In relative abundance. This tendency was also noticed for iodine of higher concentration in 3 M-HNO₃.     

Role of Effective Distance in the Fission Mechanism Study by the Double-energy Measurement for Uranium Isotopes

Fission product kinetic energies were measured by the double-energy method for thermal-neutron fission of ^235,233U and proton-induced fission of ²³⁸U at the 15.8-MeV excitation. From the obtained energy-mass correlation data, the kinetic-energy distribution was constructed from each mass bin to evaluate the first moment of the kinetic energy for a given fragment mass. The resulting kinetic energy was then converted to the effective distance between the charge centers at the moment of scission. The effective distances deduced for the proton-induced fission was concluded to be classified into two constant values, one for asymmetric and the other for symmetric mode, irrespective of the mass though an additional component was further extracted in the asymmetric mass region. This indicates that the fission takes place via two well-defined saddles, followed by the random neck rupture. On the contrary, the effective distances obtained for thermal-neutron induced fission turned out to lie along the contour line at the same level as the equilibrium deformation in the two-dimensional potential map. This strongly suggests that it is essentially a barrier-penetrating type of fission rather than the over-barrier fission.     

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%.     

超临界CO_2从含铀氧化物粉末中直接萃取铀方法研究

采用超临界CO2/硝酸-磷酸三丁酯(SC-CO2/HNO3-TBP)为萃取体系,研究各影响因素对其直接从含铀氧化物粉末中萃取铀的影响规律。结果表明:HNO3/TBP比值增大,铀萃取效率(EU)明显增大,比值为1.4时有利于萃取;比值一定时,随有机相总体积增加EU呈线性增加;萃取反应大致分为3个阶段:体系趋于稳定时的较慢阶段,稳定状态下的快速萃取阶段以及受粉末状态影响的缓慢阶段;随温度升高,EU在60℃时达到最大值91.4%;随萃取体系压力增大,EU在12 MPa时达到萃取峰值94.7%;萃取产物分离采用二级减压分离方式,在1#分离器8 MPa、2#分离器5 MPa条件下,能对萃取产物实现很好地分离。     

离子注入碳对铀表面初始氧化行为影响的研究

用俄歇电子能谱(AES)分别对高真空室中铀样品及多能量叠加离子注入碳铀样品与氧气吸附及初始氧化过程进行了研究.结果表明,在高真空室中,清洁铀表面很容易吸附氧化,发生氧化反应,当O2暴露剂量为40 L时,清洁铀表面就会形成一层UO2;离子注入碳后铀表面的抗氧化能力增强.     

金属铀与铝薄膜界面的俄歇电子能谱研究

以磁控溅射沉积方法 ,采用循环氩离子轰击镀和未循环轰击镀工艺在金属铀上制备了铝薄膜。俄歇电子能谱分析结果表明 :循环氩离子轰击镀获得的铝薄膜和铀基体的界面扩散比未循环轰击镀的显著增强 ,且界面发生化学反应 ,生成了UAl3和Al2 O3相。     

Characterization of Uranium Oxide Thin Film Prepared by Vacuum Deposition Method

Semiconductor bulk crystals and multilayer structures with controlled isotopic composition have attracted much scientific and technical interest in the past few years. Isotopic composition affects a large number of physical properties, including phonon energies and lifetimes, bandgaps, the thermal conductivity and expansion coefficient and spin-related effects. Isotope superlattices are ideal media for self-diffusion studies. In combination with neutron transmutation doping, isotope control offers a novel approach to metal-insulator transition studies. Spintronics, quantum computing and nanoparticle science are emerging fields using isotope control.     

Dissolution and Solidification of Aluminum Capsule in Production of Mo-99 by Sublimation from Neutron-Irradiated UO2

A technique, based on the correlation analysis of signal pulses in time sequence, is proposed to determine the production rate of the pseudo multi-pulse in Geiger-Mueller (GM) counter. With a multi-channel scaler initiated by a signal pulse, subsequent pulses are recorded in sequence. The production of the multi-pulse increases the counting probability immediately after the initiation. By examining the deviation of the measured probability from the ideal counting probability, the production rate and the average lag time to produce the multi-pulse can be determined. By the use of the present technique, the production rate and the average lag time were obtained for the various GM tubes. These results indicate that the consumption of the quench gas results in a significant increase in the production rate but little variation in the lag time, and that the lag time strongly depends on the tube diameter.     

Half-Lives and Gamma-Ray Energies of Short-Lived Molybdenum Isotopes

Short-lived molybdenum isotopes produced by the ²³⁶U(n, f) reaction were isolated by rapid paper electrophoretic technique. Gamma-ray spectra of the isolated molybdenum fraction were measured with high resolution Ge(Li) detector. The half-lives of ¹⁰³Mo (70±1 sec), ¹⁰⁴Mo (66±2 sec) and ¹⁰⁵Mo (30±3 and 40±5 sec) were determined by analyzing the genetic relationships of the known photopeaks of their technetium daughters. Decay analysis of observed photopeaks free from interference resulted in the detection of many new short-lived photopeaks with half-lives of 10–70 sec, and which were not relevant to technetium isotopes. These new photopeaks were ascribed to ^103-106Mo, based on examination of their decay behavior.     

Fundamental Study on Separation of Uranium from Other Metals by Solvent Extraction with Tributylphosphate or Tributylphosphine Oxide in Molten Naphthalene

The extraction behaviors of U and other metals has been studied as functions of the concentration of nitric acid or hydrochloric acid in aqueous media and as functions of the amount of naphthalene as diluents, when tributylphosphine oxide (TBPO) or tributylphosphate (TBP) was used as a extractant. Uranium could be quickly extracted over 95% from 25 cm³ of the 1 mol·dm^?3 nitric acid solution with 150 mg of TBPO and 350 mg of naphthalene at 80°C. On the other hand, about 80% of U could be extracted under better conditions, when 25 cm³ of 4 mol·dm^?3 nitric acid solution and 1cm³ of TBP and 3 g of naphthalene were used. In either case, the extraction phases could be easily separated as a solid by cooling to room temparature. Uranium could be separated from alkaline earth metals, transition metals and rare earth metals, and U was efficiently concentrated by extraction with molten naphthalene, compared with usual solvent extraction which used toluene, etc.     

Separation of Pu-238 and Pu-242 from Irradiated Am-241

Failure analysis was made on samples taken from type 304 stainless steel piping systems (core spray, unloading and feed water pipes) that had cracked in service. In the core spray pipe, large cracks including one penetrating through the wall were found in the upper half of the pipe wall, within the heat-affected zone of the weld joint between the pipe and the nozzle safe end of the reactor pressure vessel. These cracks were of intergranular type, also small transgranular cracks were found beyond the heat-affected zone. Strong correlation was established between the intergranular cracking and severe sensitization in the heat-affected zone.

In the unloading pipe, the region close to the weld joint contained cracks similar to those in the core spray pipe. The feed water pipe, in contrast to the foregoing cases, contained numerous shallow transgranular cracks both within and outside the heat-affected zone.     

Diffusion and Evaporation of Fission Products in Coated Fuel Particles

Fractional releases of ¹³³Xe, ¹⁴⁰Ba and ⁸⁹Sr from slightly-irradiated pyrolytic-carbon-coated and SiC-coated particles were measured over a temperature range of 1,200°–1,750°C. The results are analyzed mathematically in order to obtain the diffusion and evaporation coefficients relevant to PyC and SiC. The resulting expressions for the coefficient of diffusion in PyC are 2.9x10-⁷ exp(-61x10³/RT) for ¹³³Xe and 4.7x10-² exp(51x10³/RT) for ¹⁴⁰Ba. For the coefficients of evaporation of ¹⁴⁰Ba from PyC, the expression is 3.5x10³ exp(-42x10³ /RT). As for SiC, the diffusion and evaporation coefficients of these nuclides are given for a temperature of 1,750°C. A high diffusivity path for the diffusion of ¹⁴⁰Ba is postulated to explain the difference in diffusion behavior between ¹³³Xe and ¹⁴⁰Ba in PyC.     

羟基磷灰石与天然磷灰石去除铀的效果和机理研究

通过静态实验分别研究了反应时间、温度、pH值、铀初始质量浓度、磷灰石用量等条件对羟基磷灰石与天然磷灰石去除铀的影响,并根据热力学和动力学原理进行了分析。结合红外光谱、扫描电镜、X衍射等分析结果探讨了羟基磷灰石与天然磷灰石去除铀的机理。羟基磷灰石去除铀的结果表明:铀的去除量随着铀初始浓度的增大而增大;在pH=4~5的铀溶液中,羟基磷灰石用量为0.75g时,去除效果最好,铀的去除率达85%。天然磷灰石去除铀的结果表明:铀的去除率随铀初始浓度的增大而增大;在pH=3、天然磷灰石用量为1.0g时去除效果最好,铀的去除率高达80%。二者的去除反应均在120min时基本达到平衡,反应过程均既符合二级动力学方程又符合Langmuir热力学方程。红外分析结果表明,二者主要是通过磷酸根与UO2+2的络合作用去除铀。X衍射分析结果表明,羟基磷灰石基本符合纯物质的组成与构成,而天然磷灰石主要成分为Ca5H2(PO4)3F、Ca8H2(PO4)6H2O。扫描电镜对比分析显示,羟基磷灰石外观呈球状且开有孔,孔内有一含大量絮状物的空腔,羟基磷灰石在去除铀后表面变得光滑且微孔也被封闭,这是由于吸附了UO2+2使羟基磷灰石表面分子间发生了链接;而天然磷灰石去除铀前后皆呈现棱角分明的矿物外形,说明被吸附天然磷灰石上的UO2+2,形成了矿物晶体。     

Vaiidition of ORIGEN Computer Code by Measurements on Nuclear Fuels of JPDR-1

The depletion and production amounts of U, Pu, transplutonium nuclides and fission products (FPs) measured on the fuel of JPDR-1 were corrected to take account of the performance history of irradiation and cooling using the results of three-dimensional nuclear-thermo-hydrodynamic and nuclide depletion and production calculations. Except a few nuclides, the corrected values proved to agree well with the values calculated by the ORIGEN computer code. Further enhancement of calculational accuracy calls for systematic re-evaluation of neutron cross sections on the basis of neutron spectrum in nuclear fuel.     

Temperature and Concentration Dependence of Tritium Release Behavior from Solidified LiF-PbF2 Pellet Irradiated by Thermal Neutron

Thin pellets of the LiF-PbF₂ system with no porosity were irradiated by thermal neutron. Then release experiment of tritium produced in the pellets was carried out at some constant temperatures. Thus diffusion coefficients of tritium in the LiF-PbF₂ system were estimated to obtain an activation energy of diffusion, which was found to be 0.51 eV at LiF mole fraction of 0.467. It was also observed that decrease of LiF concentration in the pellet brings increase of diffusion coefficient of tritium.     

可活化稳定同位素示踪-中子活化分析法研究育龄妇女膳食铁的生物利用率

采用两种浓缩丰度分别为 61 4%和 2 3 4%的可活化稳定浓缩同位素54 Fe和58Fe作为示踪剂 ,结合中子活化分析法研究了北京地区健康育龄妇女膳食铁的生物利用率 ,并以原子吸收分光光度法测定粪便排出的总铁和以不易被吸收的镝元素监测浓缩同位素铁在体内的滞留时间以确定不同试验者的粪便收集次数等多种途径来提高分析结果的可靠性。研究结果显示 ,受试妇女膳食铁的生物利用率为(1 4 9± 3 9) %。     

Melting Temperature of Irradiated UO2 and UO2-2wt%Gd2O3 Fuel Pellets up to Burnup of about 30 GWd/tU

Melting temperature of UO₂ and UO₂-2w/oGd₂O₃ fuels irradiated in a commercial LWR were determined by a thermal arrest technique in a burnup range up to approximately 30GWd/tU.

No decrease in the melting temperature was observed on both UO₂ and UO₂-2w/oGd₂O₃ fuels with increment of burnup to 30GWd/tU. It was also found that the Gd₂O₃ addition below 2w/o has no influence on the melting temperature.     

可活化稳定同位素^70Zn示踪研究儿童膳食中锌的生物利用率

使用中子活化分析法,以国内仅有的低深缩度^70Zn（18．3％）作为可活化稳定浓缩同位素示踪剂,对我国儿童膳食中锌的生物利用率进行了研究。建立了以阴离子树脂交换法对儿童便样中的锌元素进行富集,经反应堆中了辐照、测量和放射化学分离、测定锌的生物利率方法。提出了普适于各种^70Zn浓缩度锌作为示踪剂、以^70Zn和^68Zn和^64Zn的质量丰度进行生物利率率的计算公式。克服了以^68Zn作为唯一的天然锌指示而引入的计算误差。     

DHDECMP-TBP/煤油萃取模拟高放废液中的裂片元素工艺研究

实验测定了22％ DHDECMP-42% TBP/OK从模拟高放皮液中萃取Zr、Mo、Cs等裂片元素的分配比，并在微型离心萃取器(转鼓φ10mm)串联台架上进行了从模拟高放废液中萃取裂片元素的工艺研究。采用6级萃取，2级洗涤，6级反萃，流比AF：AX：AS＝1：1．5：0．5，BF：BX＝1：1，实验考察了裂片元素的各级浓度分布。对Zr、Mo的净化系数DF分别为3．2、18；对Rb、Cs、DF、Ba，DF＞10^3。