新型氚蒸馏装置的研制及性能验证

为提高硅胶吸附法监测环境空气中氚的样品前处理效率,采用铸铝加热方式结合温度自动控制技术,研制了一种用于环境硅胶样品自动化氚水蒸馏的装置。经实验验证,该装置选择250℃的加热温度,15 min的蒸馏冷凝收集时间为硅胶吸附氚化水的最佳解析条件,此时对硅胶中氚化水的平均回收率大于99%,氚记忆效应小于0.2‰,100 g饱和硅胶样品能解析水7～10 ml,满足液闪测量要求,研制的氚水蒸馏装置较以往装置蒸馏效率提高了近3倍,能耗减少约77%。     

高压氘氚气氛下铜对氚的吸附

为了解氚在铜表面的吸附和解吸行为，对铜样品在n（D）∶n（T）=1∶1，503 K时，15 MPa下恒温8 h后，再在27 MPa恒温6 h下进行了氚的吸附，并对吸附氚的铜样品在室温下和加热到1 173 K时的解吸氚量和吸附总氚量进行了测量。结果表明，铜的吸附总氚量为31.89 MBq/cm²,解吸氚量为29.18 MBq/cm²，测量的标准差为6.49％；室温和加热条件下铜所释放的氚中，化学成分主要是HTO和HT，大部分以HT形式存在；铜的自由氚量占吸附总氚量的3.64 ％；铜的热解吸谱至少存在3个解吸峰，其解吸温度分别为650，750和1 173 K以上。     

含氚重水脱氚方法

CANDU6核电站慢化剂重水中的氚浓度随反应堆运行而增加,含氚重水以气态或者液态向反应堆厂房及环境中泄漏,将造成运行人员辐射剂量及环境污染。如何有效将慢化剂重水中的氚脱除受到国际上广泛的关注。本文调研国内外重水脱氚技术,对各种现有脱氚技术进行了系统的比较与分析,总结了各工艺的优缺点及经济性。     

CuO床氚气氧化效率测定

氚工艺中可用CuO床将氚气（HT）氧化成氚水（HTO），本实验测定了不同氧化床温度和载气流速条件下的氧化效率，以确定合适的氧化床操作工艺。     

退役核设施不锈钢中氚的分析

在自制的氚污染不锈钢的高温热解析制样系统中对氚污染不锈钢样品进行解析分析,并对影响不锈钢样品中氚回收率的解析温度、解析时间、催化氧化效率、冷凝收集效率等条件进行了优化研究。结果表明,当载气（空气）流速为60～80 L/h,解析温度和催化氧化温度为700 ℃,冷凝温度为-20 ℃,解析时间为2 h时,氚污染不锈钢样品高温制样系统的冷凝收集效率为99%,催化氧化效率为98.6%,不锈钢样品中氚的解析率＞95%。以上结果表明,该不锈钢高温热解析制样系统及其解析条件能够满足退役核设施不锈钢样品中氚的分析要求。     

高压氘氚气氛下不锈钢吸附氚的初步研究

本工作对于氚在不锈钢表面的吸附和解吸行为进行了初步研究.样品在n(D)∶n(T)=1∶1,230℃时,15 MPa下恒温8 h后,接着在27 MPa下恒温6 h的情况下进行了氚的吸附,测量了室温下和加热到1173 K时的解吸氚量和总吸附量.其结果如下:不锈钢的总吸附氚量是857.4 MBq·cm-2,不锈钢的解吸氚量是722.2 MBq·cm-2;在本实验的条件下,在室温和加热条件下,不锈钢所释放的氚中,化学成分主要是HTO和HT两部分,大部分以HT形式存在;不锈钢的自由氚占吸附总氚量的7.34%;不锈钢的热解吸谱存在三个解吸峰,其解吸温度分别为439、530和843K.     

钒床氚滞留量的测量

分别采用吸附解吸法、同位素交换法和溶解法测量钒床中的氚滞留量,并分析了这三种测量方法在本实验条件下的误差.吸附解吸法测量钒床的氚滞留量的结果如下:钒床的氚滞留份额为4.40％,当压力读数在1500～133332 Pa之间时,测量的标准差小于0.95％;同位素交换法测量钒床氚的结果如下:加热充分解吸过的钒床经多次同位素交...     

离线产氚实验中氚在线监测系统的研制

在增殖剂离线产氚实验中,如何准确实时测量回路中氚浓度和形态（HTO/HT）对于掌握产氚增殖剂氚释放行为,改进增殖剂的产氚性能非常重要。针对离线产氚回路中载气流量小、回路中气体量小以及载气为Ar等特点,基于流气式电离室原理研制了一套数字化氚浓度在线测量系统。该系统中电离室灵敏体积为50 mL,数字化仪表可自控获取、处理及显示回路中的氚浓度。测试结果表明,在Ar气氛下,在35 V左右,电离室即进入饱和区;该系统探测下限可达3.7×10 7 Bq/m 3,能满足离线产氚实验中氚在线监测的要求。     

环境大气氚的测定技术研究

文章分析环境大气中氚的来源，并描述环境大气氚浓度测量系统的组成与测量原理。研究了累积氚取样装置中催化炉的催化温度和气体流速对催化效率的影响、冷阱冷却温度对氚回收率的影响，刻度了氚电解浓集装置的回收率，在此基础上测定了某核设施周围环境大气氚浓度。结果表明：在催化温度为350℃、催化效率为98%、电解浓集装置平均回收率为54%的实验条件下，本测试方法可高灵敏度地测定特定范围内外环境低水平大气氚浓度变化，探测下限可达0.1Bq/m³。     

液态锂铅合金的氚释放行为

为了完成聚变堆液态锂铅包层鼓泡提氚系统的工程设计和建造,以金属与氢的作用理论为基础,建立了氚从液态锂铅中的动力学释放行为的数学模型.计算和分析了温度、饱和器氚分压、氦流量对解吸器顶部气相中的氚分压的影响以及氚在液态锂铅中的传质系数、解吸率和吸附率.结果表明:在633～723 K的解吸温度范围内,氚从液态锂铅到气相的整个释放过程虽然包含了氚在熔融合金气泡中的扩散与对流、氚通过与气-液界面相连合金层的扩散、在界而发生的氚原子重组多相反应、氚通过气相边界层的扩散和气相中氚的扩散与对流5个子过程,但起决定作用的是氚在合金内的扩散和气.液界面的多相反应重组,其他子过程意义不大.     

Comparison of Precious Metal Catalysts with Hydrophilic Porous Substrate for Tritium Cleanup System

Abstract

Existence of the structure water in the hydrophilic porous substrate can give various effects on tritium behavior in a tritium recovery system consisted of precious metal catalyst bed and adsorption bed because the catalytic isotope exchange reaction or the isotope exchange reaction proceeds besides oxidation reaction or adsorption reaction.

Properties of Pt catalysts loaded on various hydrophilic porous substrate are compared in this study. It is known from experiment that Pt-Alumina catalyst has the largest exchange capacity and that Pt-MS 5A catalyst has the largest total capacity because it has the effective adsorption capacity. For silica gel substrate, certain transport reaction of tritium in the substrate whose rate is rather slower than the rate of isotope exchange reaction is observed and it gives peculiar undesirable quality for silica gel to use in a tritium recovery system.     

Detrapping behavior of tritium trapped via hot atom chemical process in neutron-irradiated ternary lithium oxides

Tritium detrapping behavior in neutron-irradiated ternary lithium oxides was investigated by the comparison of the annihilation of irradiation defects with the tritium release. It was revealed that the annihilation of irradiation defects would consist of two processes; namely the fast and the slow ones. The slow annihilation process has correlation with the tritium release, indicating that E′-center or F⁺-center could act as tritium trapping site, and from its activation energy of each sample, the annihilation of E′-center and F⁺-center could be attributed to the recovery of oxygen via diffusion, triggering the tritium release. Meyer-Neldel plots of these results indicate that the slow annihilation process was governed by the formation entropy of a pair of vacancy and interstitial atom of oxygen. Therefore, the trapped tritium would be detrapped by oxygen recovery to E′-center or F⁺-center, and its kinetics would be determined by the population of oxygen vacancy under thermal equilibrium.     

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100-600 °C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300-600 °C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150 °C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.     

D+注入Li4SiO4的表面化学状态及其热解吸行为

锂陶瓷氚增殖剂的氢同位素行为是聚变堆固态产氚包层关心的重要课题。本文将3 keV D⁺注入Li₄SiO₄,采用X射线光电子能谱在线分析注氘前后材料表面的化学状态,同时采用热解吸谱（TDS）实验技术,研究注氘后Li₄SiO₄中氢同位素的热解吸行为。实验结果表明：D⁺注入会改变Li₄SiO₄表面的化学环境,产生多种辐照缺陷和化学键合状态;氘滞留量和热解行为受注氘时样品的温度影响较大,可在一定程度上预测产氚包层中氚的滞留行为。     

A thermal desorption study of the surface interaction between tritium and type 316 stainless steel

The thermal desorption of tritium from a type 316 stainless steel exposed to gaseous tritium (HT-gas) was studied. In the desorption spectra of tritium, four distinct peaks appeared at about 430 (peak HT-1), 540 (peak HTO), 750 (peak HT-II) and 970 K (peak HT-III). The activation energies for the peaks HT-I, HTO and HT-II were determined as 20 ± 4, 27 ±7 and 55 ±9 kJ/mol, respectively. It is presumed that the tritium leading to the peak HTO forms tritoxyl ion (OT^?) by combining with surface O^2? ion or by the isotopic exchange with the protium of surface OH^? ion on the sorption process. The OT ion combines with the OH^? ion or the dissolved protium into HTO at the desorption process. The tritium leading to both peaks HT-II and HT-III is sorbed and dissociates into atomic species which diffuse into the bulk, while that leading to the peak HT-I is very weakly held on the topmost surface of stainless steel in the form of a molecule or HT⁺ ion.     

氚在真空贮存容器材料中的吸附和渗透行为

研究了贮存氚靶约4 a和20 a的两个316 L不锈钢真空贮存容器(以下简称贮存容器)及其垫片材料对氚的吸附行为,并对氚在贮存容器材料中的渗透速率进行了测量和分析。结果表明,贮存容器外表面氚污染为几十Bq/cm2,不锈钢与陶瓷中吸附的氚活度均为106Bq/g;热解吸至1 273 K过程中,材料中99%的氚释放出来;在解吸出的氚中,陶瓷中的HTO比例高于不锈钢;贮存温度对氚靶贮存容器的渗氚速率有较大影响,夏季约为冬季的4倍。上述结果提示,氚在贮存容器材料内表面吸附后,一部分会向晶格扩散并滞留下来;另一部分则透过材料向外环境渗透,其中温度是影响氚向外环境渗透的主要因素之一。     

Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Release behavior of tritium from the graphite tiles used at dome top and inner dome wing in JT-60U was investigated by the thermal desorption method in dry argon, argon with oxygen and water vapor, or argon with hydrogen. It was found that approximately 20-40% of total tritium is left in graphite even after heating to the high temperature above 1000 °C in dry argon. The residual tritium could be removed by exposing the graphite tile to oxygen with water vapor or hydrogen at the high temperature above 1000 °C. The tritium retention of the dome top tile was quantified as 84-30 kBq/cm². The inner dome wing tile had a steep tritium distribution from 8 to 0.1 kBq/cm². It is observed that a measurable amount of tritium existed in the deep site of the graphite tile.     

Recovery of Hydrogen Isotopes by Pd-coated ZrNi from Inert Gas Atmosphere Containing Impurities

Effect of Pd coating on recovery of hydrogen isotopes from inert gas atmosphere was investigated for intermetallic compounds ZrNi. Powder of ZrNi was coated with Pd by electroless plating. Subsequent annealing at 1,073 K led to the formation of ZrPd at the powder surface. The durability in deuterium recovery from Ar gas was examined for the bare and the Pd-coated powders at temperatures from 373 to 573 K as well as the kinetics of deuterium absorption and desorption. The absorption and desorption of deuterium obeyed the first and the second order kinetics, respectively. The ZrPd layer on ZrNi did not alter the absorption and the desorption kinetics. The deuterium recovery by the bare powder was degraded with time owing to oxidation of powder surfaces by impurities contained in the Ar gas. The durability in deuterium recovery was markedly improved by Pd coating, and no significant degradation was observed for the Pd-coated specimen at 573 K. This difference in the durability was ascribed to the difference in the oxidation behavior between ZrNi and ZrPd. It was concluded that Pd coating significantly improves the applicability of ZrNi to tritium recovery in blanket systems.