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《中国原子能科学研究院年报》2019,(0)
<正>1核素制备研究为推进裂变~(99)Mo的生产工艺研究,继续开展LEU-Al合金靶件生产裂变~(99)Mo关键技术研究和铀箔靶件生产裂变~(99)Mo的工艺研究。为去除铅铋堆因中子辐照产生的~(210)Po,继续开展铅铋冷却剂中~(210)Po处理技术研究。为解决~(99)Mo分离新技术的关键问题,"功能性离子液体萃取分离~(99)Mo及其辐射稳定性研究"获得了基金的批准立项。 相似文献
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溶液堆在医用同位素的生产方面具有一些优势,本文对溶液堆的发展过程进行了介绍,对用于医用同位素生产的水溶液均相反应堆的技术特点、核素生产以及相关的核燃料处理问题进行了综述.溶液堆可以提取的同位素主要有99Mo, 131I, 89Sr等.在核燃料处理方面,溶剂萃取法是切实可行的方法,针对硫酸和硝酸2种溶液体系,推荐了硝酸体系的φ=30% TBP流程.溶液堆运行1~2年左右,冷却3~5个月进行后处理,放射性浓度大于99%的裂变和腐蚀产物被去除,铀的回收率大于99.5%,回收的铀可以回堆继续应用,形成一个快速处理循环.在后处理设备方面,小型化的核用离心萃取器及过滤设备是最好的选择. 相似文献
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与裂变型~(99)Mo-~(99m)Tc发生器相比,凝胶型~(99)Mo-~(99m)Tc发生器制备~(99m)Tc具有工艺简单、产生的放射性废物容易处理、对环境影响小等优点。本文主要论述了凝胶型~(99)Mo-~(99m)Tc发生器与裂变型~(99)Mo-~(99m)Tc发生器的区别,堆照生产~(99)Mo原料和凝胶材料的研究进展,凝胶结构以及凝胶组分等多种条件因素对凝胶型~(99)Mo-~(99m)Tc发生器性能的影响等,并对低比活度~(99)Mo生产~(99m)Tc的研究进展进行综述。 相似文献
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~(99)Mo是一种重要的医用放射性同位素。采用低浓铀(LEU)靶件生产裂变~(99)Mo是发展趋势。本工作进行了电沉积UO_2靶件制备、靶件溶解以及99Mo化学分离等工艺研究,确定了电沉积LEU UO_2靶件制备医用裂变99Mo的工艺流程。研究表明,于不锈钢管内壁上电沉积UO_2,在p H=7、电流0.5~2 m A/cm2、温度75~90℃、镀液中U浓度5 mg/mL条件下,经过约210 h电沉积,不锈钢管内壁上UO_2沉积层质量达到42 mg/cm~2;采用6 mol/L HNO_3溶解UO_2镀层。采用α-安息香肟沉淀法实现~(99)Mo与大量裂变产物的初步分离,采用阴离子交换法与活性炭色层法联用实现99Mo的纯化;纯化后的99Mo溶液中,杂质~(131)I、~(90)Sr、~(95)Zr、~(103)Ru、~(238)U活度与~(99)Mo活度比值分别为4.47×10~(-6)%、7.40×10~(-7)%、8.67×10~(-7)%、2.57×10~(-6)%、1.69×10~(-14)%,均小于《欧洲药典》规定值,满足医用要求。本工作建立了电沉积LEU UO_2靶件生产高纯医用裂变99Mo的工艺流程,为今后采用LEU技术生产医用裂变99Mo,进而实现其自主规模化生产打下了基础。 相似文献
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《中国原子能科学研究院年报》2017,(0)
正当前,医用~(99) Mo主要从~(235)U裂变产物中提取。~(235)U裂变产物含有100多种核素,其中~(131)I(T_(1/2)=8.02d)裂变产额为2.84%,相对于~(99) Mo(T_(1/2)=2.747d),其半衰期较长。因而,在医用裂变~(99) Mo生产工艺研究中,~(131)I的去除是关键技术之一。中国原子能科学研究院正在进行低浓缩铀(LEU)生产裂变~(99) Mo的工艺研究,~(99) Mo的化学分离工艺为:用HNO_3溶解辐照过的铀靶件,首先采用α-安息香肟(α-BO)沉淀法从铀靶模拟溶解液中分离~(99) Mo,实现~(99) Mo与其他裂变杂质的分 相似文献
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为适应当前全球对放射性诊断核素99m Tc需求量不断增长的需要,阿尔及利亚比林核研究中心对多用途重水反应堆(Multi-purposes Heavy Water Research Reactor,MHWRR)实施升级改造,提出建立不停堆连续辐照生产裂变钼的能力需求。在对裂变钼靶件堆内辐照堆芯物理计算分析、热工水力计算分析、靶件出堆过程热工计算分析以及裂变99Mo产额计算等分析研究的基础上,结合反应堆设施原有限制条件,创新地提出了采用短时间临时停堆方式的技术方案,既能实现阿方产量目标,又能满足辐照安全要求。方案得到了阿方认可,工程实施后的初步调试结果表明:理论计算值与实验值符合较好,在无参考可借鉴实例的情况下,提出的辐照技术方案和工艺流程是合理可行的。 相似文献
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The status and the problems of world 99Mo production are presented. A comparative analysis is made of reactor methods of 99Mo production. It is noted that the currently used technologies and research reactors are not satisfying the growing demand
in medicine for this isotope. It is underscored that the role of alternative production technologies has grown. In the development
of new 99Mo production technologies, the experimental results obtained on the basis of research program conducted on the MSRE reactor
with molten-salt fluoride fuel have been analyzed. The analysis revealed a special behavior of certain fission products including
99Mo: they leave the melt spontaneously and enter the gas phase. The authors hypothesize that highly volatile fluorides of the
indicated products are formed in the melt; this explains the effect indicated. The effect is used as a basis to propose a
new reactor method of producing fissionproduced 99Mo. Concrete examples of a way to implement the new method of producing fission-produced 99Mo using molten-salt fluoride nuclear fuel are presented. 相似文献
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一、前言 对~(235)U热中子裂变已经作了充分的研究,裂变产物的产额作了广泛的测量,并对实验数据进行了多次编评。但是对于其它单能中子诱发~(235)U裂变研究还远远不够,尤其是keV能区的中子更是如此。J.G.Cuninghame等测量了130-1700keV中子诱发~(235)U裂变中一些核素的产额。但是低于130keV中子诱发的裂变研究,文献中未见过报道。为了研究产额随中子能 相似文献
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Conclusions This comparison of two ways of making99Mo shows that there is good scope for making it in any area. If there is a thermal reactor having a high flux, one can make99Mo from98Mo, and in that case, even irradiation in a high flux makes it favorable to use highly enriched98Mo and unblocked targets, which raises the specific activity and thus increases the working life in the99mTc generator.If there is a thermal reactor with low flux or if there is a fast reactor it is best to make99Mo from uranium fission products. The target can be highly enriched235U, which if necessary can be reused, or low-enriched uranium.There are no essential constraints on making99Mo, and the production is mainly based on technological tasks.Translated from Atomnaya Énergiya, Vol. 67, No. 2, pp. 104–108, August, 1989. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(10):551-556
The fission rate in the core of the Japan Research Reactor 4 (JRR-4) was determined by a method based on radiochemical analysis of 99Mo formed in the U samples irradiated in the reactor core. The contribution of epithermal neutron fission to the total fission rate was evaluated from the Cd ratio for U fission. The contribution was several percent. For comparison, the thermal neutron flux also was measured, by Au-foil activation. The fission rate determined from the U samples agreed well with the Au-foil data, except at positions in the peripheral region of the reactor core. 相似文献
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A physical model has been developed to describe the coolant activity behaviour of 99Tc, during constant and reactor shutdown operations. This analysis accounts for the fission production of technetium and molybdenum, in which their chemical form and volatility is determined by a thermodynamic treatment using Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix and vaporization from the fuel-grain surface. Based on several in-reactor tests with defective fuel elements, and as supported by the thermodynamic analysis, the model accounts for the washout of molybdenum from the defective fuel on reactor shutdown. The model also considers the recoil release of both 99Mo and 99Tc from uranium contamination, as well as a corrosion source due to activation of 98Mo. The model has provided an estimate of the activity ratio 99Tc/137Cs in the ion-exchange columns of the Darlington Nuclear Generating Station, i.e., 6 × 10−6 (following ∼200 days of steady reactor operation) and 4 × 10−6 (with reactor shutdown). These results are consistent with that measured by the Battelle Pacific Northwest Laboratories with a mixed-bed resin-sampling device installed in a number of Pressurized Water Reactor and Boiling Water Reactor plants. 相似文献
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运用NJOY99程序,以微观评价库ENDF/B-Ⅶ.0为基础,开发了适用于快堆研究设计的175群中子、42群光子的多群截面数据库MUSE-F1.0。采用权重谱thermal--1/e--fast reactor-fission+fusion及勒让德P6近似。采用ANISN程序,从临界计算及屏蔽计算两方面对该库进行了较全面的检验;屏蔽检验涉及裂变堆、聚变堆、加速器等装置屏蔽材料所常用的相关核素截面数据的检验。检验结果表明:MUSE-F1.0在临界计算及屏蔽计算方面具有较高的精确度和较强的适用性,可满足快堆设计研究方面的应用要求。 相似文献
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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。 相似文献