铝土矿中铀、钍、钾分布的地球化学特征以及这些特征在寻找铝土矿中的应用

一书的第59-66页中详细地介绍了苏联铝土矿床中铀、钍、钾分布的规律性,以及在这些铝土矿中及其围岩中铀、针、钾含量的不同。叙述了铝土矿床经凤化后及其所形成的次生分散晕中铀、社、钾的性状。列举     

BP型中子、γ混合埸吸收剂量探测噐系统

本文介绍了自制的BP型中子、γ射线混合场吸收剂量探测器系统。对该探测器的基本性能进行了研究。给出了球形、指形外推形电离室的主要特性。对于15MeV中子场和~(60)Coγ射线场所有修正因子都接近于1。示出了微剂量测量用的正比计数管的主要参数以及在中子、γ混合场中线能y谱的测量结果。描述了由圆盘形TE电离室和GMγ射线剂量计组成的监测系统并给出此监测系统的稳定性。在电信号处理系统中使用了VFC变换器。还谈到了在中子、γ混合场中测量y分布所得各段谱的联结技术.     

对供方的评价和选择

指出了核电站建设中采购物项和服务的重要性、法规和标准中对供方的评价和选择的规定和要求；论述了评价的目的、内容、方式和活动；介绍了秦山重水堆核电站主合同中关于对供方的资格评定。     

Purex流程中界面物的形成和机理研究

论述了Purex流程中萃取界面物的成因和影响因素,包括不同浓度的磷酸二丁酯、磷酸一丁酯、锆、铀和硝酸以及辐照剂量不同的磷酸三丁酯正十二烷溶液。观察了锆、钼、铌和锝在形成界面物过程中的化学行为,在低于2mol/L的硝酸溶液中,钼会进入界面物中。给出了界面物组分的主要元素磷/锆和锆/钼之摩尔比;分析研究了红外光谱;深入讨论了界面物的形成机理;导出了主要反应方程式。结果指出:萃取过程中由于生成了     

钋的自沉积研究

在各种浓度的盐酸溶液中进行了钋在铜、镍、银和金上的自沉积实验,在金上的自沉积是在含硫脲的盐酸溶液中进行的。讨论了对沉积速度公式中速度常数K的影响因素,计算了K值。     

三代项目核岛设计计划编制经验反馈

在参与EPR三代核电项目的建设过程中,以先分解、了解局部、再整体结合的观察方法,将其核岛设计计划编制、管理体系与CPR1000的设计计划编制、管理体系做了比较。尤其重点对EPR核岛设计计划中先分块编制,再通过逻辑关系联系成一个整体的做法作了相关研究。其中,重点关注了EPR设计计划中EPS、WBS分级与CPR1000计划中相关内容的异同,同时还就不同分块计划之间接口计划的作用作了相关阐述。其后又详细介绍了核岛土建、系统、布置三大类设计计划的具体编制方法,最后就计划中关键节点时间的确定作了相关推理论述,以达到工程经验反馈的目的。     

田湾核电站维修质量控制管理

介绍了田湾核电站核岛维修过程中质量控制(QC)管理的组织机构、QC检查的作用、QC检查员的分级及能力要求、维修QC检查的内容。提出了田湾核电站核岛维修过程中质量控制管理检查中存在的问题及改进措施。     

核设施环境与三废数据管理系统的研制

研制了民用核设施的环境、辐射防护和三废数据管理软件。详细讲述了在软件的研制过程中，利用ＶｉｓｕａｌＦｏｘＰｒｏ为开发工具，运用面向对象的设计方法和拖放、动画、ＯＬＥ等ＷＩＮＤＯＷＳ高级功能，解决了在数据管理中遇到的难点，阐述了整个工作中技术上的创新和该管理系统的优点。     

铀和锆在镉中的分布

对Ｕ－１０Ｚｒ合金在４７５、５００℃下Ｕ、Ｚｒ在液态Ｃｄ中的分布状态进行了研究。用ＪＣＸＡ－７３３电子显微分析仪测定了金属Ｃｄ中Ｕ、Ｚｒ的浓度，采用元素特征Ｘ射线面扫描法分析了Ｕ、Ｚｒ在金属Ｃｄ中的分布状态。结果表明：Ｕ、Ｚｒ在金属Ｃｄ中分布是均匀的；溶解了Ｕ－１０Ｚｒ合金的金属Ｃｄ中存在少量微小的颗粒（析出物），此类析出物中含有金属Ｃｄ原料中已存在的杂质元素Ｋ。     

浙江省水体中天然放射性水平调查

本文介绍了浙江省水体中天然放射性核索水平调查方案和分析工作的质量保证措施,给出了江河、湖泊、水库、海、泉和自来水中 U、Th、~(226)Ra 和~(40)K 的含量。调查结果表明,尚未发现浙江省水体中有明显的放射性污染;但石煤、磷矿和铀矿的开采使钱塘江下游支流中天然放射性核素浓度有增高的趋势。     

Calibrating a gas chromatograph to measure tritium using calorimetry

The availability of calibrated gas mixtures in a tritium plant is an important precondition for correct analysis of unknown gas samples because a quantitative analysis with a gas chromatograph (GC) requires frequent calibration checks. To calibrate a GC for tritium, certified gas mixtures with different tritium concentrations are necessary. These mixtures can be produced directly in a laboratory. In this case an independent method is needed to determine the tritium concentration in the mixture with high accuracy. Calorimetry is the method of choice because its accuracy is better than 1% and it is non-destructive method.Using the closed tritium loop of the Tritium Laboratory Karlsruhe (TLK), different mixtures containing tritium in the range 1–100% have been produced and measured with one of the in-house calorimeters. After that the GC of the CAPER facility at TLK has been calibrated several times with an uncertainty of a few percents.     

Tritium diffusivity in crystal grain of Li2TiO3 and tritium release behavior under several purge gas conditions

It has been reported by the present authors that behavior of tritium release from solid breeder grain is consisted of diffusion in grain, tritium transfer at surface layer and surface reactions on grain surface such as adsorption or isotope exchange reactions. Tritium release curves estimated using the tritium release model gave good agreement with observed tritium release curves from Li₄SiO₄, Li₂ZrO₃ or LiAlO₂.

Tritium release behavior from Li₂TiO₃ under humid purge gas, dry purge gas and dry purge gas with hydrogen conditions is discussed in this study, tritium release curves using the release model that we proposed previously give a good agreements with experimental tritium release curves. Tritium effective diffusivity in the crystal grain of Li₂TiO₃ is also estimated in this study using a curve-fitting method applied to the release curves obtained under the humid purge gas condition. It is discussed that change of color of Li₂TiO₃ surface under hydrogen purge gas condition is observed and this phenomenon might affect tritium release behavior from Li₂TiO₃.     

氚在橡胶中的渗透

本文介绍了氚在橡胶中的渗透原理,并对10多种橡胶材料的氚渗透性能进行了测量。测量结果表明,天然橡胶的氚渗透系数最大,74氯丁橡胶最小;氚气(HT)对天然橡胶的渗透系数是74氯丁橡胶的1.58倍;厚度为1mm的74氯丁橡胶和天然橡胶膜片达到稳态渗透所需延迟时间分别为72和4小时,相差18倍。     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

氚气接触吸收严重程度评估—一起氚内照射事件的启示

本文以一起氚气内照射事件中工作人员的尿氚测定结果为依据,试对氚气接触吸收的严重程度进行一些定量分析。结果表明,皮肤跟被氚气污染了的金属表面接触时所吸收的氚与相同面积的皮肤在相同时间内从与污染金属表面处于吸附解吸平衡状态的含氚空气中浸没吸收的氚的比值约为5×10~3;双手臂与污染金属表面接触所吸收的氚与呼吸道和全身皮肤从空气中浸没吸收的氚之和的比值约为2×10~2。从而强调了氚气接触吸收的严重性。     

Overview of the tritium system of Ignitor

Among the recent design activities of the Ignitor program, the analysis of the tritium system has been carried out with the aim to describe the main equipments and the operations needed for supplying the deuterium–tritium mixtures and recovering the plasma exhaust.

In fact, the tritium system of Ignitor provides for injecting deuterium–tritium mixtures into the vacuum chamber in order to sustain the fusion reaction: furthermore, it generally manages and controls the tritium and the tritiated materials of the machine fuel cycle. Main functions consist of tritium storage and delivery, tritium injection, tritium recovery from plasma exhaust, treatment of the tritiated wastes, detritiation of the contaminated atmospheres, tritium analysis and accountability.

In this work an analysis of the designed tritium system of Ignitor is summarized.     

聚变实验增殖堆氚系统设计研究(英文)

介绍了核工业西南物理研究院聚变实验增殖堆工程概要设计(FEB-E)中的氚系统设计研究。第一部分介绍包层氚增殖区的划分、几何尺寸、装料特征和用蒙特卡罗程序计算得到的液态锂中的氚浓度分布;第二部分描述根据聚变堆氚物理基础构造的氚循环系统,共分成 10 个子系统及它们之间氚的流程图。运用研制的程序SWITRIM 计算了各个子系统中的氚投料量随时间的变化,满功率运行一年后各个子系统中的氚投料量。研究结果表明起动 143 MW 聚变功率 FEB-E 堆所需要的初始氚投料量大约为 319 g。第三部分对不同的运行状态下的氚泄漏问题进行了分析。潜在的氚泄漏危险可能来自于偏滤器系统从等离子体中抽出的气体。得到的结论是提高FEB-E 堆芯等离子体的燃耗份额从而减少氚的通过量对降低氚的泄漏危险是重要的。     

Tritium management and safety issues in ITER and DEMO breeding blankets

Safe, reliable and efficient tritium management in the breeder blanket faces unique technological challenges. Beside the tritium recovery efficiency in the tritium extraction and coolant purification systems, the tritium tracking accuracy between the inner and outer fuel cycle shall also be demonstrated. Furthermore, it is self-evident that safe handling and confinement of tritium need to be stringently assured to evolve fusion as a reliable technique. The present paper gives an overview of tritium management in breeder blankets. After a short introduction into the tritium fuel cycle and blanket basics, open tritium issues are discussed, thereby focusing on tritium extraction from blanket, coolant detritiation and tritium analytics and accountancy, necessary for accurate and reliable processing as well as for book-keeping.     

Tritium retention to the first wall of JT-60U

For economic and safety reasons, tritium (T) accumulation on plasma facing wall (PFW) of fusion reactor is strictly limited. In this study, T inventory in the graphite tiles used at the first wall of JT-60U was measured by a full combustion method. It was found that T was only retained near plasma facing surfaces sides of the tiles and the amount of retained T increased from <10¹¹ to <10¹³ T atoms/cm² with increasing the exposed discharge period of the tiles. Integrating the T retention with the total surface area of the outer first wall tiles, the fraction of retained T in that area was estimated to be 13% of the total T production. It was confirmed that these retained T were part of the energetic T produced by DD reactions without being replaced by HH discharges. Based on the retained fraction, the annual amount of T retention in the outer first wall of a demo-size reactor was calculated to be 360 g/burn-year at maximum. Even the value would be much less in the reactor, the accumulation of this kind of inventory could have significant contribution to the total T inventory.