压阳离子交换-α-HMBA淋洗色层分离Am,Cm和Cf

Ferguson等报道,用α-羟基-α-甲基丁酸(α-HMBA)作阳离子交换分离超钚元素的淋洗剂,可将Am,Cm的分离因子提高到1,46。我们测定了α-HMBA与Am,Cm和Cf的络合稳定常数,与陆兆达等人的结果相近,Am和Cm与α-HMBA的络合稳定常数之比均明显高于α-羟基异丁酸(α-HIBA)的相应值。因此可以认为,α-HMBA是目前已研究过的用于阳离子交换分离超钚元素的络合剂中较为优良的一种。     

阳离子交换-DTPA-AHIB络合淋洗法分离Am,Cm和Pm

在阳离子交换淋洗色层中,α-羟基异丁酸(AHIB)用于分离超钚元素已达30年之久。然而,用AHIB分离和纯化Am和Cm时,Pm在Am和Cm之间被洗脱,影响Am,Cm产品的纯度。Елесин等用氨基多乙酸-AHIB溶液分离Am,Cm和Eu,Pm时发现,在所研     

水-醇混合介质中离子交换分离超钚元素——Ⅱ.CH_3OH-LiNO_3-HNO_3-DTPA-阴离子交换树脂体系

序言 离子交换色层法是超钚元素分离的有效方法。不论是阳离子交换或阴离子交换都有过报道。我们曾研究过CH_3OH-HNO_3-阴离子交换树脂分离体系和CH_3OH-HNO_3-NH_4NO_3-DTPA-阴离子交换树脂分离体系。其Am,Cm分离因子前者为2.3,后者为3.5。本文的主要目的在于寻找进一步提高Am,Cm分离因子及探寻Am,Cm与裂片的分离。     

α-羟基异丁酸阳离子交换色层法分离Am、Cm、Cf和Y

描述了以α-羟基异丁酸为淋洗剂的阳离子交换色层法分离Am,Cm,Cf和Y的实验程序,得到了令人满意的分离结果。为了把感兴趣的超钚元素淋洗在所需要的位置上,研究了α-羟基异丁酸的PH和浓度对淋洗峰位置的影响。此外,报道了实验测得的Am、Cf、Y对Cm的分离因子。     

微克量级SUP>173,174/SUP>Lu的

本工作以α-羟基异丁酸(α-HIBA)为淋洗剂,用聚苯乙烯磺酸型阳离子树脂(002×8型,粒度为40～50 μm)交换柱,从克量级的天然Yb辐照靶中分离出了微克量的173,174Lu.研究了温度、载体量等因素对分离效果的影响,并分别研究了阳离子交换法、浓HClO4-HNO3破坏法、氢氧化物沉淀法、HDEHP萃取法和HDEHP萃取色层法从介质中去除α-HIBA的实验条件.最终在0.07 mol/L α-HIBA淋洗液(pH=6.2),流速5 mL/min,柱温55 ℃的条件下,经φ20 mm×500 mm交换柱分离2次,成功从550 mg辐照样品中得到0.8 μg 173,174Lu,收率为98%,对Yb的去污因子大于104.     

加压淋洗色谱法从∧147Nd中分离∧147Pm的研究

研究了用加压分离交换法从堆照产物∧147Nd中分离∧147Pm。首先用加压排代色谱法确定了堆照产物∧147Nd中的主要杂质成分,继而用加压淋洗色谱法以α-羟基异丁酸（α-HIBA）-抗坏血酸（Vc)为淋洗剂;从∧147Nd中分离∧147Pm,避免了主要杂质∧152,154Eu对产品∧147Pm的污染,研究结果表明,当分离条件为：c(α-HIBA)=0.30～0.40mol/L、c(Vc)=0.05～0.10mol/L、pH=4.00～4.20、线性流速为8～12cm/min时,分离因子β(Nd/Pm）=3.00,β(Eu/Nd)=2.61。     

居里级~(242)Cm的制备新工艺

本文研究了一个提取居里级~(242)Cm的化学流程,采用TBP萃取色层,除裂片元素,CH_3OH-HNO_3-阴离子交换树脂体系分离Am和Cm,产品中~(241)Am的α放射性与~(242)Cm的α放射性之比达到3×10~(-4)。此流程具有步骤少,操作方便和产品介质简单等优点。     

镅(Ⅲ)、锔(Ⅲ)阳离子交换分离的优良淋洗剂——α-羟基α-甲基丁酸

本文在相同的实验条件下,在阳离子交换柱上比较了α-羟基α-甲基丁酸(HMBA)和α-羟基异丁酸(HIBA)为淋洗剂分离Am(Ⅲ),Cm(Ⅲ)的性能。实验测得对HIBA体系,Am(Ⅲ)-Cm(Ⅲ)的分离因子为1.37,与文献值基本一致;而对HMBA体系,其分离因子为1.71,此值与我们测定的络合常数相吻合,从而确证以HMBA淋洗分离Am(Ⅲ)-Cm(Ⅲ)远较HIBA优良。为了使HMBA成为实用的淋洗剂,我们还研究了影响分离效果的各种因素。     

季铵盐萃取分离镅、锔

一、引言镅、锔分离是超钚元素化学中的难题之一。将Am氧化到五价然后分离的方法只能用于工艺粗分离,且操作繁杂。目前常用的α羟基异丁酸阳离子交换法,分离因子仅为1.5。甲醇-硝酸阴离子交换法,分离因子也只有2.4。近年一些文章报道了用胺类萃取法或萃取色层法分离Am、Cm,优点是胺类萃取剂耐辐照、分离因子也高。Horwitz等研究了用季铵盐从含盐析剂的硝酸中萃取三价     

综合提取锕系和镧系元素工艺流程中Am(Cm)的测定

本文研究了从强放废液中综合提取锕系和镧系元素过程中Am(Cm)的测定。采用0.05 MPMBP-0.025MTOP/环己烷溶液协同萃取Am(Cm),以三倍于有机相体积的二甲苯稀释荷载有机相,再用0.05M DTPA-1.0M乳酸(pH=3,下同)溶液定量反萃Am(Cm)。在0.2MHDEHP/煤油和0.05MDTPA-1.0M乳酸萃取体系中将Am(Cm)与剩余镧系元素分离。最后制源,测定Am(Cm)α放射性强度。方法对样品中Am(Cm)的回收率为(97.0±5.4)%     

二(2,4,4三甲基戊基)-二硫代膦酸萃取分离Am和Cm的研究

研究了二（2,4,4三甲基戊基）-二硫代膦酸（HBTMPDTP）对Am与Cm的萃取和分离,单级萃取分离因数约为3。pH值、离子强度、萃取剂浓度、温度等因素对分离因子影响不大。8级萃取、3级洗涤的多级逆流萃取实验表明：HBTMPDTP能够使Am与Cm得到有效分离。萃取实验的计算值和实验值符合很好。根据串级计算给出了HBTMPDTP萃取分离压水堆废液中镅和锔的推荐流程,采用9级萃取、4级洗涤的萃取分离流程,镅的萃取率为99．92％,放化纯度达到99．99％,质量纯度达到99．82%;锔的萃余率为94．90％,放化纯度达到99．90％,质量纯度达到97．68％,镅中锔的分离因数为20,锔中镅的分离因数为1186。可以满足Am与Cm的分离-嬗变要求。     

镅锔分离研究进展

乏燃料后处理产生的高放废液中Am和Cm是长期释热的主要来源,将它们分离出来并进一步进行分离和处置,对高放废物的长期安全处理处置具有重要意义。另外,超钚元素生产涉及Am和Cm材料的获取以及辐照后靶件中Am和Cm的化学分离。因此Am、Cm的分离一直是锕系元素化学与材料研究的重要领域之一。但是Am、Cm之间的分离相当困难,水溶液中Am、Cm基本均以正三价离子形式存在,化学性质非常相似。早期的离子交换法分离因子低,近年来主要研究将Am(Ⅲ)氧化到高价态实现分离,或通过Am、Cm与配体的亲和力差异、不同配体组合产生“推拉效应”以提高分离因子。本文综述了相关研究现状,概述了主要流程研发情况,并展望了该领域的研究趋势。     

水-醇介质中离子交换分离超钚元素——Ⅰ.CH_3OH-NH_4NO_3-HNO_3-DTPA-阴离子交换树脂体系

本文研究了Am和Cm在CH_3OH-NH_4NO_3-HNO_3-DTPA溶液和阴离子交换树脂之间的分配系数。结果表明,DTPA的存在对Am和Cm的分离有显著的改善,二者的分离系数可达3.5-3.7。 以适当的淋洗液进行柱分离实验,达到了Am,Cm,Cf三个元素的分离。     

Homogeneous Transmutation of Am, Cm, and Np in the Core of a BREST Reactor

The effect of homogeneous transmutation of Am, Cm, and Np in fuel on the radiation characteristics of the fuel is studied. It is shown that the holding period of the spent fuel influences the composition and mass of Am, Cm, and Np and the fuel quality. For homogeneous transmutation, increasing the Am, Cm, and Np fraction in the fresh fuel up to 5% has virtually no effect on the basic neutron-physical characteristics and safety of the reactor. The effect of adding to the main actinides particular elements from Am, Cm, and Np on the radiation characteristics of fresh and spent fuel with homogeneous transmutation is studied. 4 tables, 6 references.     

Np,Am, Cm utilization in a nuclear power reactor

The utilization (transmutation using the released energy) of Np, Am, and Cm in a specialized VVER-type reactor operating power 1000 MW(e) is studied by computational means. Five successive reactor runs, where the fuel elements contain Np, Am, Cm, and enriched uranium, are studied. It is shown that Np, Am, and Cm produced by several VVER-1000 reactors can be utilized in one run (900 days) of a specialized reactor without any technological and structural changes. Up to 4% of the reactor power comes from the fission of these radionuclides.     

Np,Am, Cm transmutation efficiency parameters taking account of neutron reaction rates

Np, Am, and Cm transmutation efficiency parameters, taking account of the neutron reaction rates on all transuranium isotopes in all transformation chains initiated by introducing Np, Am, and Cm, are proposed for specialized transmutation facilities. These parameters can be helpful when comparing the efficiencies of different types of transmutation facilities and analyzing possible ways to increase the Np, Am, and Cm transmutation efficiency. The efficacy of including the parameters in the definition of transmutation efficiency is shown for a specialized transmutation facility. The possibility of evaluating the efficiency of Np, Am, and Cm transmutation in fast power reactors is examined taking account of the proposed parameters.     

Long-Lived-Radionuclide Transmutation Scenarios

The results of multigroup calculations of continuous irradiation of Np, Am, and Cm in VVÉR-1000, PHWR-880, Superphoenix-1200, BREST-1000, and ÉLYaU-800 reactors are used to compare transmutation efficiency. The sources of continuous replenishment for the transmuters were Np, Am, and Cm extracted after a 3-yr holding period from the VVÉR and Superphoenix spent fuel. It is shown that the most effective transmuter is a subcritical liquid-fuel ÉLYaU system with an average thermal-neutron flux in the blanket 2·10¹⁵ sec^–1·cm^–2. For solid-fuel reactors, the continuous-irradiation model makes it possible to describe approximately the multiple transmutation regime. In the foreseeable future, one-time transmutation of Np, Am, and Cm in a solid-fuel reactor followed by storage in a long-term storage facility is feasible. The results of different computational variants for such regimes show that for transmutation in 10 yr in PHWR the radiotoxicity of Np, Am, and Cm accumulated in long-term storage reaches an equilibrium in no longer than 100 yr.     

Specific activity of243Am and243Cm in the fuel of the 4th power-generating unit of the Chernobyl nuclear power plant

The activity ratios²⁴¹Am/²⁴¹Am.²⁴³Cm/²⁴⁴Cm, and²⁴²Cm/²⁴⁴Cm in core samples taken at the industrial site of the object “Cover” were measured. The content of²⁴³Am and²⁴³Cm in the fuel in the 4th power-generating unit of the Chernobyl nuclear power plant was estimated on the basis of an analysis of these data. The experimental data were compared with the theoretical calculations. 2 figures, 9 references. Scientific Center “Institute of Nuclear Studies,” Ukrainian National Academy of Sciences. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 327–329, November, 1999.     

Characterization of radiation fields and dose assessment from fuels manufacturing for advanced fuel cycles

The purpose of this study was to examine radiological hazards introduced to workers from the fabrication of fuels with minor actinides (MA) and reprocessed uranium (RepU) and determine the feasibility of using shielded gloveboxes instead of remotely controlled operations in hot cells. Of particular concern is the increase in photon source term from the daughter products in RepU and the mixed neutron and photon fields introduced by MA recycle. In the interest of keeping the glovebox worker's radiation dose as low as reasonable achievable (ALARA), dose rates were calculated for various typical and bounding fuel compositions for light water reactors (LWR) and fast reactor (FR) mixed oxide (MOX) fuels with and without MA. The impact of varying the separation efficiency of americium (Am) and curium (Cm) was examined because current separation processes in reprocessing and recycling do not allow for the complete separation of Am from Cm. The additional Cm will cause a significant increase in the neutron source term. The sensitivity of the fuels to aging time was also examined by decaying the recycled feedstocks from 6 months to 3 years to simulate the effect of delays in reuse of recycled materials. The highest photon and neutron sources were used to calculate the additional shielding requirements that would be needed for fuel fabrication. Through insight gained from this study, it can be concluded that a standard glovebox with one quarter inch stainless steel walls can be used to fabricate fuels with Am with little to no additional shielding. The introduction of small amounts of Cm in the fuels will require the fuel fabrication to be preformed remotely in hot cells. Thus, stressing the importance of developing methods to increase the separation efficiency of Am from Cm.     

Heterogeneous Transmutation of Am, Cm, and Np in a BREST Core

The effect of heterogeneous (in individual units) transmutation of Am, Cm, and Np on the radiation characteristics of fuel is examined. For heterogeneous transmutation of Am, Cm, and Np in individual fuel elements containing nitrides, the radiation characteristics and energy release increase substantially compared with fresh homogeneous fuel elements. Heterogeneous transmutation is dangerous from the standpoint of nonproliferation of fissioning materials because of the low critical mass of the main nuclides – ²³⁹Np and americium isotopes. 2 tables, 4 references.