酰胺荚醚对Am（Ⅲ）和Eu（Ⅲ）的萃取行为研究 Ⅱ萃取势力？…

以４０％辛醇／煤油为稀释剂,研究了３种荚醚：Ｎ,Ｎ,Ｎ＾’,Ｎ＾’－四丁基－３－戊二酰胺（ＴＢＯＰＤＡ）、Ｎ,Ｎ,Ｎ＾’,Ｎ＾’－四异丁基－３－氧－戊二酰胺（ＴｉＢＯＰＤＡ）和Ｎ,Ｎ,Ｎ＾’,Ｎ＾’－四丁基－３,６－二氧－辛二酰胺（ＴＢＤＯＯＤＡ）在硝酯介质中对Ａｍ（Ⅲ）和Ｅｕ（Ⅲ）的萃取热力学。ＴＢＯＰＤＡ、ＴｉＢＯＰＤＡ和ＴＢＤＯＯＤＡ萃取镅的反应焓变分别为：－８０．５４、－８１．９９和－７５     

铀氢锆堆物理计算及燃料管理软件包

建立了一套铀氢锆堆物计算软件包，首先考虑氢化锆中的热化特殊性，按ＷＭＳ格式制作 了氢化锆 氢的６９群群常数并入ＷＩＭＳ－Ｄ／４数据库中，形成了ＷＩＭＳ－Ｎ１库和ＷＩＭＳ－Ｎ２库；应用ＷＩＭＳ－Ｎ２库和国际通用的ＷＩＭＳ－Ｄ／４程序包计算了铀氢锆堆各类栅元的群常数，应用差分程序ＣＩＴＡＴＩＯＮ和六角形节块和ＳＩＸＴＵＳ进行扩散计算，同时在ＳＩＸＴＵＳ－２程序的基础上编制了燃料管理程序和ＸＰＲ－ＩＣＦ     

新萃取剂N-酰化吡咯烷的合成及萃取U(Ⅵ)的性能

以吡咯烷为二级胺,合成了Ｎ－辛酰吡咯烷（ＯＰＯＤ）、Ｎ－葵酰吡咯烷（ＤＰＯＤ）、Ｎ－月桂酰吡咯烷（ＤＯＰＯＤ）等三种新型双取代酰胺类－Ｎ－酰化吡咯烷类（Ｎ－ａｃｙｌａｙｒｒｏｌｉｄｉｎｅ ＡＰＯＤ）萃取剂,利用红外光谱,核磁共振谱对其结构进行了表征,并对该类萃取铀的性能进行了研究。     

硝酸铀酰与TDD配合物合成及性质研究

在氯仿－甲醇混合剂中，合成了Ｎ，Ｎ，Ｎ＇Ｎ＇－四苯基－３，６－二氧杂辛二酰胺（ＴＤＤ）与硝酸铀酰的固体配合物，其分子式为［ＵＯ２（ＮＯ３）２（ＴＤＤ）］。通过元素分析、红外光谱、热分析、＾１ＨＮＭＲ谱和摩尔电导率对该配合物进行了表征。     

硝酸铀酰与TDD配合物的合成及性质研究

在氯仿－甲醇混合溶剂中，合成了Ｎ，Ｎ，Ｎ’Ｎ’－四苯基－３，６－二氧杂辛二酰胺（ＴＤＤ）与硝酸铀酰的固体配合物，其分子式为［ＵＯ＿２（ＮＯ＿３）＿２（ＴＤＤ）］。通过元素分析、红外光谱、热分析、￣１ＨＮＭＲ谱和摩尔电导率对该配合物进行了表征。     

中国评价核数据库第二版(CENDL-2)

中国评价核数据库第二版（ＣＥＮＤＬ－２），包含从Ｈ到中的５４个重要核的全套评价中子核数据，中子能区为１０－５２０ＭｅＶ。它是在ＣＥＮＤＬ－１基础上，进行更新评价、扩展能区、扩充核素和数据种类；按国际标准格式建立计算机化核数据库；改进和发展理论计算、评价处理的有关方法和技术，研制配套相应的计算程序和基准检验程序系统；评价分析和理论计算细致，部分核的评价有自己的特色，较国际上先进的评价核数据库有所改进，有些评价数据还被这些先进库所采用。国际原子能机构已把ＣＥＮＤＬ－２作为国际上最先进的５个主要评价核数据库之一向世界各国发行，提供使用。     

铍中的反应率测量和分析

用４种不同阈能的活化箔测量和比较了两种铍系统中的绝对活化反应率。铍厚分别为５．３和８．４ｃｍ,,活化箔为Ｉｎ,Ｚｎ,Ａｌ和Ｃｕ,实验总２为５％－７％,分析和比较了实验结果与用１－ＤＡＮＩＳＮ程序和ＥＮＤＦ－Ｂ／Ⅵ库数据的计算结果,两者间的偏差均小于７％。     

核聚变实验装置HT-7U及大厅活化分析

使用一维ＳＮ离散坐标法输运程序ＡＮＩＳＮ、活化计算程序ＡＦＤＫＲ并使用一维球几何模型对ＨＴ－７Ｕ装置米要部件及混凝土屏蔽墙的活化进行了计算和分析,给出了中子能谱、γ能谱、大厅内外剂量率空间分布及放射性水平的时间分布,对ＨＴ－７Ｕ装置的主要中子学参数及周围居民接收剂量水平给出了定量的分析。     

心肌显像剂99mTcN-BMPBDA的制备及动物体内分布研究

为了寻找新的具有亲心肌性质的＾９９ｍＴｃＮ药物,合成了一种新的Ｎ２Ｓ２配体：Ｎ,Ｎ＇－二（２－巯基丙基）－１,２－苯二胺（ＢＭＰＢＤＡ）。以ＳｎＣｌ２为还原剂,Ｈ２ＮＮＨ－（Ｃ＝Ｓ）－ＳＣＨ３为Ｎ＾３－给予体,通过交换反应制备了放化纯大于９０％的＾９９ｍＴｃＮ－ＢＭＰＢＤＡ。探讨了ｐＨ值对＾９９ｍＴｃＮ－ＢＭＰＢＤＡ放化纯的影响,并对其小鼠生物分布进行了研究。结果表明,＾９９ｍＴｃＮ－ＢＭＰＢＤＡ     

酰胺荚醚对Tc(Ⅶ)的萃取行为研究

研究了３种荚醚Ｎ,Ｎ,Ｎ’Ｎ’－四丁基－３－氧－戊二酰胺（ＴＢＯＰＤＡ）、Ｎ,Ｎ,Ｎ’Ｎ’－四异丁基－３－氧－戊二酰胺（ＴｉＢＯＰＤＡ）和Ｎ,Ｎ,Ｎ’Ｎ’－四丁基－３,６－二氧－辛二酰胺（ＴＢＤＯＯＤＡ）在ＨＮＯ３介质中对ＴｃＯ＾－４的萃取行为,稀释剂为４０％辛醇－煤油。萃取过程为一放热反应。萃取反应式为Ｈ＾＋＋ＴｃＯ＾－４＋ｎＳ＝ＨＴｃＯ４．ｉＳ（式中Ｓ表示萃取剂）。在酸度为０．５、２．０、４．     

核聚变实验装置HT-7U停机辐射剂量率三维计算与分析

在基于三维蒙特卡罗方法的聚变装置停机剂量率计算方法“严格两步法”(R2S)的基础上,首先建立了核聚变托卡马克实验装置HT-7U三维精确模型,然后对HT-7U各种D-D放电模式下的停机剂量率进行了详细的三维计算与分析,从而为装置实验方案及实验维修人员的安全操作规程的制定提供了理论基础,也对装置的辐射防护问题具有参考价值。     

Comparison of the Feasible Schemes for HT-7U Assembly Simulation

The whole superconducting HT-7U Tokamak is a high-cost and large-scale compli-cated device.The assembly requiremem of HT-7U device is arduous and strict.At present,therehave been no guiding principle for the assembly of the device,but assembly simulation can help theengineer plan and make decision by an intuitional and visual way before its actual assembly.The     

Design of the Cryostat for HT—7U superconducting Tokamak

The cryostat of HT-7U tokamak is a large vacuum vessel surrounding the entire basic machine with a cylindrical shell,a dished top and a flat bottom.The main function of HT-7U cryostat is to provide a thermal barrier between an ambient temperature test hall and a liquid helium-cooled superconducting magnet.The loads applied to the cryostat are from sources of vacuum pressure,dead weight,seismic events and electromagnetic forces originated by eddy currents.It also provides feed-through penetrations for all the conecting elements inside and outside the cryostat.The main material selected for the cryostat is stainless steel 304L.The structural analyses including buckling for the cryostat vessel under the plasma operation condition have been carried out by using a finite element code.Stress analysis results show that the maximum stress intensity was below the allowable value.In this paper,the structural analyses and design of HT-7U cryostat are emphasized.     

Recent Progress of HT-7U Superconducting Tokamak

HT-7U is a superconducting tokamak. which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak.The engineering design of the device has been optimized. The R&D program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.     

The renewed HT-7 plasma control system based on real-time Linux cluster

In order to improve the synchronization, flexibility and expansibility of the plasma control on HT-7, a new plasma control system (HT-7 PCS) was constructed. The HT-7 PCS was based on a real-time Linux cluster with a well-defined, robust and flexible software infrastructure which was adapted from DIII-D PCS. In this paper, the hardware structure and system customization details for HT-7 PCS are reported. The plasma position and current control, plasma density control and off-normal event detection, which were realized in separated systems originally, have been integrated and implemented in such HT-7 PCS. All these control algorithms have been successfully validated in the last several HT-7 experiment campaigns. Good control performance has been achieved and the experiment results are discussed in the paper.     

Radiation Losses in Ohmically Heated Plasma in the HT-7 Tokamak

Total radiation losses in ohmically heated plasma have been investigated in the HT-7 tokamak after different wall conditionings. Discharges after siliconization and boronization have been studied by using bolomitric measurements. Not only low-Z impurities but also high-Z impurities are reduced obviously after wall conditioning. The fraction of the total radiation loss to the ohmic input increases with increasing line-averaged central electron density. But it decreases with increasing the plasma current IP after wall conditionings. It is slightly lower for a boronized wall than that for a siliconized wall.     

Evaluation of The Thermal Performance of Multi—Element Doped Graphite under Steady—State High Heat Flux

Multi-element doped graphite,GBST1308 has been developed as a plasma facing material(PFM) for high heat flux components of the HT-7U device.The thermal performance of the material under steady-state(SS) high heat flux was evaluated under actively cooling conditions,the specimens were mechanically joined to copper heat sink with supercarbon sheet as a compliant layer between the interfaces.The experiments have been performed in a facility of ACT (actively cooling test stand) with a 100kW electron gun in order to test the suitability and the loading limit of such materials.The surface temperature and bulk temperature distribtuion of the specimens were investigated.The experimental results are very encouraging that when heat flux is not more than 6 MW/m^2,the surface temperature of GBST1308 is less than 1000℃,which is the lowest,compared with IG-430U and even with CX-2002U(CFC),The primary results indicate that the mechanically-joined material system by such a proper design as thin tile.Super compliant layer,GBST as PFM and copper-alloy heat sink,can be used as divertor plater for HT-7U in the first phase.     

HT-7 超导托卡马克上辐射剂量测量与分析

HT-7超导托卡马克是我国正在运行中的中型准稳态磁约束聚变实验研究装置,运行时产生一个复杂的瞬发混合辐射场。本文介绍了该装置上中子辐射剂量率的测量方法以及在欧姆、波辅助加热、长脉冲等放电模式下的剂量率特点,并且给出了X与7射线的总辐射剂量测量方法和分析结果。在严格遵守安全操作规程的前提下,目前的等离子体物理实验对人员产生的总辐射剂量低于国家辐射安全标准的规定。     

HT-7超导托卡马克上聚变中子产额和中子注量率分布计算

利用所编制的三维托卡马克中子计算程序,计算了HT-7装置上中子产额随等离子体中心离子温度与密度的变化.同时计算了中子注量率及其在空间不同位置与方位角的分布特性.考虑了离子温度与密度分布的形状因子不同所带来的误差.计算结果为中子诊断和中子辐射防护等提供了依据.     

二维中子-光子输运计算程序DOT3.5在核聚变实验装置HT-7U辐射屏蔽设计中的应用

利用二维中子－光子输运程序ＤＯＴ３．５计算和分析ＨＴ－７Ｕ在Ｄ－Ｄ放电运物时实验大厅内外的中子、光子通量及剂量当量率的空间分布,并评估大厅顶部不同防护材料的屏蔽效果,为辐射防护设计提供参考。