固体氚增殖剂的制备及性能综述

聚变或聚变一裂变混合堆中通常采用含锂的材料作为氚增殖剂.概述并比较了氚增殖剂的几种制备方法,就几种氚增殖剂的热学、化学、机械性能、辐照行为和氚释放等进行了综合评述,简单介绍了氚增殖剂在聚变堆中的作用和发展趋势,为我国实验包层模块(TBM)产氚包层中氚增殖剂的深入研究奠定了基础.     

液态金属包层中氚提取技术的研究进展

为了完成ITER液态金属包层氚提取系统（TES）的设计,针对液态包层材料——液态锂、液态锂铅、盐溶液和氟化物盐中提取氚的方法进行了调研。调研结果表明：钇粒子床法和溶盐萃取法适用于液态锂中氚的提取;膜渗透法、鼓泡器法和钠回路附设冷阱法对液态锂铅中的氚提取是很有希望的,尽管氚扩散通过锂铅合金边界层的速率缓慢,增大了氚提取的技术难度,但鼓泡器法具有更高的气一液间接触面积,自90年代以来技术发展最为成熟;氦气清洗和真空萃取法已成功运用于氟化物盐包层中氚的提取。尽管任何一种氚提取方法都有工程化的现实前景,但为解决氚提取系统中杂质影响和各种相互作用的问题,开展材料和基础化学方面的广泛研究是必要的。     

在线产氘回路及其在混合堆包层研究中的应用

介绍了我国一条反应堆在线产氘回路，说明了回路的组成及其主要技术指标，叙述了回路的运行概况和放氘实验的结论，讨论了在线产氘回路在聚变－裂变混合堆包层产氘研究中的应用及前景。     

RAFM钢中氚氦行为的研究进展

低活化铁素体/马氏体(RAFM)钢是未来磁约束核聚变堆产氚包层推荐的主体结构材料,将在ITER实验包层模块中进行考核。氚及氚衰变氦-3在RAFM钢中的驻留行为将影响所产生氚的有效利用并导致材料的结构损伤,因此RAFM钢的涉氚相容性是其最终应用于聚变堆结构材料之前必须考核的问题之一。总结了国内外对RAFM钢的氚氦相容性研究进展情况,提出了一些要解决的关键问题和将来的研究方向。     

高气密性芯片级原子气室的制备研究

氚是涉氚场所中辐射防护监测的重点核素，对涉氚场所中的气态氚浓度进行有效监测具有重要的意义。根据GB/T 7165.5-2008第5部分要求，氚活度监测校准应使用与被测对象成分一致的参考源，即标准氚气。解决该难题需要研究并建立标准氚气的制备方法和定值方法，确保氚气活度量值源头的有效性。基于稀释法研制了气载氚标准源，通过长度补偿法进行氚活度浓度定值，并进行了不确定度的分析和评定。氚气标准源的研制填补了国内该领域的空白，为氚监测仪的量值溯源提供了重要支撑。     

氚在硅中的无载体加载行为

硅是供电(光)器件制作中重要的换能功能材料,为了选择国内采用放射性同位素氚研制辐射伏特效应核电池的加栽方法,从多孔硅的制备方法出发,详细介绍了国外开展的氚在多孔单晶硅、无定型硅上的无载体加载工作,并结合其实验结果,理论分析了氚在硅表面的化学键合方式与释放行为.分析结果表明:为了得到较高的含氚量,在采用多孔单晶硅加载氚时,不仅要提高反应体系中氚的含量,还应提高硅的表面孔密度;在用无定型硅加载氚时,应控制好氚的沉积条件;新制作的氚化硅核电池在额外放电几天后才有稳定的电输出.     

低温热色谱分离氢同位素

叙述了用“６０１”碳分子筛作托卡马克废气排出气的收集、纯化和低温热色谱的进样单元，以５Ａ分子筛柱在液氮（ＬＮ２）温度下吸附氕、氘，氚，随后把柱温回升至室温（２０℃～３０℃），为了提高分离效率、缩短分离时间和分离柱总长度，在两根分离柱间添加了ＨＤ、ＨＴ等歧化平衡器，经过四根分离柱组成的级联分离柱后氕、氘、氚获得了清晰分离。该装置的单次进样量为５标准升，日处理量为１２０～１８０标准升，流程气体总回收率＞９７％，氘丰度＞９９．９％时，氘、氚损失量（不计加热３００℃活化再生分离柱回收的氘、氚）＜５％，除氕效率＞９８％。     

反应堆含氚重水提氚关键技术研究进展

基于建立的年处理１０t重水的组合电解催化交换唱色谱分离（CECE-GC）实验系统,就含氘轻水提氘 演示实验及利用含氚轻水进行含氚重水提氚模拟运行做了介绍。结果表明,２４０h含氚轻水的连续运行,CECE 系统整体浓缩倍数约为４,电解池氕、氚分离因子约为１０;８m３桙d色谱分离系统运行２３h,可将１０.５m３料气的 ９０％中的氘贫化１０００倍;氚储存系统运转正常。CECE唱GC实验系统的建立,为重水提氚技术的进一步工程化提 供了研究平台。     

LaNi4.25A10.75床氚滞留量的测量

分别采用直接测量法、同位素交换法和溶解法测量LaNi4.25A10.75床中的氚滞留量，并分析了这3种测量方法在本实验条件下的误差。直接测量法测量LaNi4.25A10.75床的氚滞留量的结果如下：LaNi4.25A10.75床的氚滞留量为3．65％（（0．097±0．00092）mmol T2／g LaNi4.25A10.75），当压力读数在1．5-133.332kPa之间时，其PVT测量法的标准差小于0．95％；同位素交换法测量LaNi4.25A10.75床氚的结果如下：加热充分解吸过的LaNi4.25A10.75床经多次同位素交换后，其交换效率仅为1．92％，即不到2％的氚被氘气载带出来，其同位素交换法测量的标准差为7．35％。溶解法能够定量地测量LaNi4.25A10.75床中滞留的氚，其溶解法测量的标准差为6．49％。     

高分子材料在氚工艺和氚防护中的应用

介绍和分析了高分子材料在基本氚工艺设备、惯性约束核聚变聚合物靶丸材料和氚个人防护中的研究和应用现状.并就建立统一的高分子材料氚渗透测试体系和进行高分子材料的各种改性研究以提高抗氚渗透性能等方面提出了建议.     

Indirect tritium determination by an original 3He ingrowth method using a standard helium leak detector mass spectrometer

A traditional helium leak detector mass spectrometer is applied in analytical chemistry. We report its straightforward use for the precise quantification of 3He and 4He concentrations at a sub-ppb level. Repetitive external calibrations with certified gases demonstrate long-term stability, reproducibility of 3% at the ppb range, and linearity versus helium concentration over 5 orders of magnitude. The mass spectrometric determination of 3He takes into account a H3+ contribution and has a 3He detection limit of 2 ppb. Since 4He induces an attenuation effect of the mass 3 signal, this apparatus is only suitable for the measurement of 3He/4He ratio greater than 10 Ra (atmospheric ratio Ra = 1.38 x 10(-6)). This mass spectrometric technique has found novel application in the determination of tritium concentrations with an original 3He ingrowth method. Here, samples are first purified cryogenically onto activated charcoal to eliminate tritium interfering during the 3He mass spectrometric determination step. Helium is also preconcentrated, and 3He is routinely determined around 0.05 ppb with a 20% relative uncertainty. This 3He measurement technique has been successfully applied in the field of tritiated waste drums as a nondestructive method. The total amount of tritium present in these 200-L drums can be determined as low as approximately 1 GBq.     

中子管充氚排气排风系统设计与使用

根据中子管生产的需要，设计一套充氚排气排风系统，经过长时间的使用效果相当好，提高了中子管生产效率和成品质量。     

Investigation of the hydrogen multilayered target H/T−D2 and muonic X-ray yields in ion implantation

This paper extends applications of the multilayered solid target H/T−D₂, which is kept at 3 K. The time evolutions of muonic tritium atoms (μt) are obtained, by taking into account μt production rate at different places of deuterium material. The apparatus H/T−D₂ can be used for checking nuclear properties of implanted ions, which take part at muon transfer. Electromagnetic X-rays are generated by muon atomic transitions. The muonic X-ray transition energies are strongly affected by the size of nuclei. Here, a solid hydrogen-tritium (H/T) with a ≈1 mm thick is used for μt production. For ion implantation, the required amount of deuterium material is determined to be about 3.2 μm. Moreover, the muonic X-ray yields are estimated and compared with those of the arrangement H/T−D₂. While the present target requires argon ion beam intensity nearly a factor of 2 times smaller; gives a relatively higher X-ray yield (15% enhancement per hour) at the energy 644 keV with the detection efficiency of ≈1%.     

Dose reconstruction for an occupational cohort at the Savannah River nuclear facility: evaluation of a hybrid method

The Savannah River Site (SRS) is the only nuclear facility in the United States that produces tritium, a radioactive isotope of hydrogen. The purpose of the study was to derive annual tritium dose estimates for SRS employees through the development of a job-exposure matrix. The proposed method is unique in that along with qualitative information on job, area and time of employment, it utilises recorded annual whole-body dose measures, when available, in order to estimate doses from tritium intakes of the monitored workers. Using information from 75 253 dose measures for the period 1954-1978, the average proportion of the whole-body dose that was due to tritium intake was calculated; these proportions were allowed to vary by job, area and time period. This information was used to assign tritium dose levels for 43 590 employment-years. The collective estimated tritium dose was 4319 mSv compared with the total known tritium dose of 17 382 mSv. The correlation (R2) of estimated tritium dose with known tritium dose was 0.68.     

Isotope Exchange of Azidothymidine with Tritium

The following modes of isotope exchange of azidothymidine (3'-azido-3'-deoxythymidine) with tritium were studied: solid- and liquid-phase isotope exchange with gaseous tritium and isotope exchange in solution with tritium water. Catalytic reactions of azidothymidine with gaseous tritium in solution result in virtually complete reduction of the azido group to amino group. This reduction also occurs in the course of solid-phase catalytic hydrogenation; the yield of 3'-amino-3'-deoxythymidine ranges from 20 to 70%. The molar radioactivity of tritium-labeled azidothymidine prepared by solid-phase catalytic isotope exchange with gaseous tritium and by isotope exchange in solution with tritium water does not exceed 0.5 Ci mmol^{- 1}.     

Distribution of tritium in water vapour and precipitation around Wolsung nuclear power plant

The distribution of tritium in water vapour and precipitation with discharge of tritiated water vapour and meteorological factors was studied around the Wolsung nuclear power plant (NPP) site during the period 2004-2008. The tritium concentrations in atmospheric water vapour and precipitation had a temporal variation with relatively high values in the early summer. Spatial distribution of tritium concentrations was affected by various factors such as distance from the NPP site, wind direction, tritium discharge into the atmosphere and atmospheric dispersion factor. The annual mean concentrations of atmospheric HTO and precipitation were correlated with the amount of gaseous tritium released from the Wolsung NPP. The tritium concentrations in precipitation decrease exponentially with an increase of the distance from the Wolsung NPP site.     

Blood tritium level as an estimate of soft tissue dose

To test the blood tritium level as an estimate of free water tritium (FWT) and organically bound tritium (OBT) in human soft tissues, a tritiated amino acid mixture (TAA) and tritiated water (HTO) were injected or orally administered to mice. Relatively high tritium retention in the blood was found after TAA injection. This observation was explained by a low rate of clearance of non-volatile tritium components. The dose estimated for the soft tissues after TAA injection was larger than that estimated on the basis of tritium concentration in urine by a factor of 1.8-2.3. In both HTO and TAA injections, the blood tritium level gave a conservative estimate of the dose to soft tissues including active bone marrow. The accumulated dose to soft tissues after oral administration tended to be lower than that for intraperitoneal injection.     

Tritium labeling of bioorganic compounds by isotope exchange

Procedures for preparing labeled compounds by liquid- and solid-phase methods using gaseous tritium and by isotope exchange with tritium water are considered as different manifestations of a common complex of processes occurring in the presence of tritium, a substrate, and a catalyst. The studies performed allow purposeful optimization of the conditions of tritium labeling of practically any biologically active substance, which makes possible more detailed investigation of the functioning of living objects.