铀及某些裂变产物的组分离

本工作是为测定铀的裂变产额而建立的组分离程序。在确定了铀靶的溶解方法和同位素交换条件下,研究了铀及其裂变产物在HNO_3-阴离子交换树脂和HCl-阴离子交换树脂上的吸附和解吸行为,选择并确定了这些元素在该体系中的吸附和解吸条件的基础上,建立了HNO_3-阴离子交换和HCl-阴离子交换为主的组分离程序。整个程序可将克量铀、几百毫克铝(包壳材料)和裂变产物分成十一组,各元素的回收率为60—95%。     

从混合裂变产物中分离铪

进行了从短期冷却的混合裂变产物中分离铪的研究。测定了在不同Zr,Hf载体量时的分离因子α和分离度R_3值,推荐了一个分离放化纯~(181)Hf的流程。方法简单,对裂变产物的去污因子为10~8～10_9,对~(95)Zr的分离因子为～10_6,化学产率为60～70％。     

放射化学分离裂变产物中^79Se的新流程

因加速器质谱法（ＡＭＳ）测定＾７９Ｓｅ半衰期的需要，建立了一个从裂变产物中放化分离＾７９Ｓｅ的新流程。流程以硝基苯萃取－二氧化硫沉淀为主要步骤，避免了经典的ＳｅＢｒ４蒸馏法带来的＾７９Ｂｒ对＾７９Ｓｅ的同量素污染。流程的化学回收率大于６０％，对所要分离的各种放射性核素的去污能力满足要求。采用液闪方法测量＾７９Ｓｅ的放射性活度，并对影响活度测量的各种因素进行了详细的研究。用本流程获得的＾７９Ｓｅ可成     

钚及其裂变产物钯 、银 、镉 、锡 、锑 、锆的系统分离方法

描述了钚及其6种裂变产物钯、银、镉、锡、锑、锆的系统分离方法：在强碱性阴离子交换树脂柱上将盐酸介质的辐照靶溶解液中的这些元素分为5组,然后再针对各组目标元素进行分离和纯化,可简便快速地从同一份靶溶解液中分离以上7种元素。采用辐照铀靶对分离方法进行了验证,结果表明,分离流程对6种裂变产物的化学回收率均大于70%,对γ谱仪测量干扰的主要核素去污因子均大于1.0×10³,可满足²³⁹Pu裂变谷区核素裂变产额测量对化学分离的要求。     

裂变产物衰变数据库

一、引言裂变产物核数据包括裂变产物产额、裂变产物衰变数据和裂变产物的中子截面数据。裂变产物核数据在反应堆方面主要用于计算衰变热。停堆后由放射性核素衰变而释放出的能量谓衰变热。衰变热的正确计算对控制动力堆的安全性有重要意义。如果冷却不     

钚及其裂变产物钯、银、镉、锡、锑、锆的系统分离方法

描述了钚及其6种裂变产物钯、银、镉、锡、锑、锆的系统分离方法：在强碱性阴离子交换树脂柱上将盐酸介质的辐照靶溶解液中的这些元素分为5组，然后再针对各组目标元素进行分离和纯化，可简便快速地从同一份靶溶解液中分离以上7种元素。采用辐照铀靶对分离方法进行了验证，结果表明，分离流程对6种裂变产物的化学回收率均大于70%，对γ谱仪测量干扰的主要核素去污因子均大于1.0×10³，可满足²³⁹Pu裂变谷区核素裂变产额测量对化学分离的要求。     

气-固分离前后裂变产物样品的γ能谱模拟

针对裂变产物气-固分离前后样品测量的预先评估需求,建立了相应的γ能谱模拟方法。采用修改独立产额数据库的方法实现了气-固分离前后裂变产物的活度计算,根据HPGe探测器测量γ能谱的原理推导了γ能峰面积计算公式,进而结合蒙特卡罗模拟软件和Power Builder10.0编程工具编制了模拟程序。利用该程序分别模拟了裂变产物无损测量和气体裂变产物测量能谱,结果与实测能谱基本相符。     

裂变产物中^138Cs的放化分离

随着科学技术的发展,对衰变数据的准确度要求越来越高。为了获得138Cs更精确的衰变数据,首先需要制备出放化纯的138Cs样品。 138Cs样品必须从新的裂变产物中分离制备。为消除同位素的干扰,本工作以衰变链中同位素核及其前驱核寿命的差异为依据,采用“两步延迟分离法”,即先等裂     

从裂变产物中提取135Cs的研究

为了制备适用于热中子截面测量用的^135Cs,建立了用无机离子交换剂磷酸锆和阴离子交换树脂从裂变产物中分离无载体放化纯^135Cs的方法。研究了在用磷酸锆色层柱分离Cs时,温度对收率的影响。实验结果表明：用磷酸锆色层柱从裂变产物中分离Cs时,Cs的化学回收率为90．0％,对锆、铌、铈、钌等的去污因子为10^2;结合阴离子交换树脂分离,U的去污因子达10^4。分离出的^135Cs为放化纯,可供^135Cs热中子反应截面测量用。     

Separation of Antimony-125 in Fission Products

Most of the known radioactive nuclides of antimony produced by neutron irradiation of uranium have fission yields below 1% and have half-lives below 60 days. An exception is ¹²⁵Sb with a half-life of 2.7 yr, which raise its relative importance among the fission products with lapse of time after irradiation, and after 1 yr of cooling, its radioactivity is no longer negligible. This circumstance has led to its being separated from such sources as fall-out. No studies have so far been reported on using the nitrate system for this separation, though it is utilized in the reprocessing of spent fuel and in the dissolution of uranium samples. The present work describes a method of separating ¹²⁵Sb from fission products with use made of silica gel—nitric acid system, and an example of its application to the separation of ¹²⁵Sb from the spent fuel of JPDR-1. The fuel was irradiated from Oct. 1963 to Sep. 1969. The amount of ¹²⁵Sb measured after separation was (1.7± O.19)×10^?1Ci/gU at June 1972.     

干馏法从铀的裂变产物中分离131I

¹³¹I是一种重要的医用放射性同位素,但因湿法分离技术上的缺陷,使得从铀裂变产物中获取¹³¹I的工艺具有环境污染严重、提取效率低的缺点。因铀裂变产物中¹³¹I的产额较高,为拓展¹³¹I的获取途径,提高铀裂变产物的利用效率,开展铀裂变产物中¹³¹I分离的新工艺研究十分必要。与传统湿法分离工艺不同,本工作采用了干馏法进行铀裂变产物中¹³¹I的分离。为了得到高的¹³¹I分离效率,将分离过程分为低温粉化、高温干馏和中低温保温三个阶段,并研究高温干馏阶段温度对131I分离效率的影响。实验发现：当干馏温度高于950 ℃时,¹³¹I的分离效率≥98%。此外,研究结果还表明,在该干馏温度下,碘和¹⁰³Ru 均可挥发出铀靶片,但产物收集液中却仅含有碘。为了解释这一现象,对碘的分离过程进行分析,结合实验结果和理论计算,推测挥发物中碘和¹⁰³Ru分离的原因为：¹⁰³Ru与氧反应生成挥发性RuO₄,从铀的裂变产物挥发出;因加热管内温度较高,RuO₄在迁移过程中发生了分解,生成RuO₂沉积在加热管内部。因此,利用干馏法从铀的裂变产物中分离¹³¹I时,为了得到放化纯度高的碘产品,不仅要合理规划分离过程,还需科学设计加热管的长度。     

裂变产物中138Cs的分离

为了获得更精确的138Cs衰变数据,需要制备出放化纯的138Cs样品.以"两步延迟分离法"为基础,将抽气法与碘铋酸铯沉淀、硅钨酸铯沉淀法相结合,建立了从裂变产物中分离放化纯138Cs的分离流程.其化学回收率达(74±1) %,对主要γ核素的去污因子大于103,操作时间在60 min以内.     

从混合裂变产物中放化分离132I

依据混合裂变产物中碘及其母体碲的同位素的半衰期设计分离¹³²I的流程。该流程的主要步骤为浓HBr蒸发和CCl₄萃取。实验研究了浓HBr蒸发对碘的去污效果；在硝酸介质中，用含I₂的CCl₄作为萃取剂，研究了HNO₃浓度、水相中KI含量和有机相CCl₄中I₂含量对¹³²I萃取率的影响，测定了含SO₂水溶液对¹³²I的反萃率。用设计的推荐流程获得了放化纯的¹³²I，其中含有的¹³¹I的活度为¹³²I的1.3％，分离流程全程对¹³²I的化学回收率约为60％，流程对主要γ核素的去污因子大于10³。     

Sequential Ion-Exchange Separation of Heavy Elements and Selected Fission Products for Burnup Measurement

A sequential ion-exchange separation method was developed for use in burnup measurements of nuclear fuels. Group separation by anion-exchange resin column with hydrochloric acid solutions containing small amounts of nitric acid and hydrochloric acid was followed by various cation and anion- exchange processes. The heavy elements, such as U, Np and Pu, and some fission products selected as burnup monitors, such as Cs, Mo and Nd, could be sequentially and quantitatively separated from a sample taken from spent fuel. The recovery of these elements through the separation processes were examined. The sampling ratio of an aliquot in reference to the whole fuel specimen was determined by adding as sampling monitor a known amount of Cu to the sample during dissolution. The validity of the ion-exchange separation technique for routine analysis for burnup measurements is also discussed.     

Gadolinium Isotope Separation by Cation Exchange Chromatography

Displacement chromatographies of Gd adsorption band in cation exchange resin were performed to observe the isotope effects in the Gd ion exchange processes involving complex forming reagents. The heavy isotope of ¹⁶⁰Gd was found to be enriched at the front boundary of Gd adsorption band and the lighter isotopes of 1MGd, ¹⁵⁶Gd and ¹⁵⁷Gd were enriched at the rear boundary in both cases of 20.1m elution with EDTA and 14 m elution with malic acid, as predicted in the theoretical relations. Observed separation coefficients are 4.9×10^?5, 4.0×10^?5 and 2.5×10^?5for isotopie pairs of ^{156 160}Gd, ¹⁵⁸Gd and ¹⁶⁰Gd, respectively, in the case of EDTA elution. In the case of malic acid elution, smaller separation coefficients were observed as 1.8×10^?5, 1.6 ⁵O^?5 and 0.92×10^?5 for isotopie pairs of ^{156 160}Gd, ¹⁵⁷Gd and^{158 160} respectively.     

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%.