CITP-Ⅱ产氚辐照装置的设计研究

在线产氚辐照装置是CITP-Ⅱ的核心部件,是研究增殖剂材料产氚及释氚试验的关键设备。本文介绍了CITP-Ⅱ产氚辐照装置的基本结构,实现了增殖剂材料的在线换料;对装置的产氚量、自屏因子、中子注量率等物理参数进行了计算;对装置的固气两相流进行了研究,估算了装置的流场特性;对增殖剂发热率、热点、温度分布梯度、极限温度、不均匀因子等热工参数进行了分析;设计了非线性的电加热器,对增殖剂的不均匀发热进行了补偿;阐述了间气及载气的成份压力对平衡温度的影响等;确定了装置的运行参数。本研究得到的关键参数及变化规律,为CITP-Ⅱ在线产氚辐照装置的结构优化及安全分析提供了依据,也可为增殖剂材料的产氚、释氚试验研究提供借鉴。     

CITP-Ⅱ产氚辐照装置的热工研究

CITP-Ⅱ在线产氚辐照装置是CITP-Ⅱ的核心部件,是研究增殖剂材料产氚及释氚试验的关键设备。对产氚辐照装置的增殖剂发热率、热点、温度分布梯度、极限温度、不均匀因子以及CITP-Ⅱ在线产氚辐照装置间隙气体及载气的成份、压力对增殖剂平衡温度的影响等进行研究,得到关键的热工参数及变化规律。     

在线产氚辐照装置物理参数模拟

在线产氚回路对我国氚增殖模块（TBM）增殖剂候选材料的考核、氚增殖剂材料的在线释放规律研究具有重要意义,辐照装置是在线产氚回路的关键部件。本工作采用MCNP程序模拟在线产氚辐照装置在堆内辐照时的物理参数,计算结果如下：自屏因子为0.430,等效反应截面为1.09×10^-22cm²,每日产量为2.8×10¹⁰Bq,总发热功率为8.2kW。模拟计算结果为该装置的设计提供了必需的数据支持。     

堆内辐照氚增殖剂球床有效热导率分析

氘-氚聚变反应堆中,固态氚增殖剂包层能不断为聚变反应提供氚核素,是实现聚变反应堆商用的关键技术之一。由锂陶瓷小球堆积形成的球床形式的固态氚增殖剂包层具有比表面积大、产氚效率高等优点,是我国重点发展的氚增殖剂包层形式。氚增殖剂球床须能支撑在堆内辐照时的高温环境,这就要求氚增殖剂球床有较好的导热特性。球床的有效热导率在球床设计和辐照过程中的安全分析十分重要,因此在中国先进研究堆（CARR）开展了氚增殖剂球床在堆内辐照环境下的有效热导率测量实验。根据MCNP计算得出的球床发热功率,结合实验测量的球床温度分布反推得到氚增殖剂球床的有效热导率,并与广泛应用于球床有效热导率计算的改进型ZBS模型计算结果以及堆外实验结果进行对比分析,理论值与实验值能较好吻合。     

堆内辐照氚增殖剂球床有效热导率分析

氘-氚聚变反应堆中,固态氚增殖剂包层能不断为聚变反应提供氚核素,是实现聚变反应堆商用的关键技术之一。由锂陶瓷小球堆积形成的球床形式的固态氚增殖剂包层具有比表面积大、产氚效率高等优点,是我国重点发展的氚增殖剂包层形式。氚增殖剂球床须能支撑在堆内辐照时的高温环境,这就要求氚增殖剂球床有较好的导热特性。球床的有效热导率在球床设计和辐照过程中的安全分析十分重要,因此在中国先进研究堆(CARR)开展了氚增殖剂球床在堆内辐照环境下的有效热导率测量实验。根据MCNP计算得出的球床发热功率,结合实验测量的球床温度分布反推得到氚增殖剂球床的有效热导率,并与广泛应用于球床有效热导率计算的改进型ZBS模型计算结果以及堆外实验结果进行对比分析,理论值与实验值能较好吻合。     

中国聚变堆材料研究

概述了中国在聚变堆材料研究方面的活动。现已研制的或正在研制的有多种结构材料、氚增殖材料和面对等离子体材料。在辐照效应、相容性、等离子体与材料相互作用、等离子体破裂期间的热冲击、氚的产生、释放与渗透以及中子在铍和铅中的倍增等方面,都已进行了系统的研究工作。文中给出部分实验结果。     

中国聚变堆材料研究(英文)

概述了中国在聚变堆材料研究方面的活动。现已研制的或正在研制的有多种结构材料、氚增殖材料和面对等离子体材料。在辐照效应、相容性、等离子体与材料相互作用、等离子体破裂期间的热冲击、氚的产生、释放与渗透以及中子在铍和铅中的倍增等方面,都已进行了系统的研究工作。文中给出部分实验结果。     

CITP-Ⅱ产氚辐照装置的换料流场研究

介绍CITP-Ⅱ产氚辐照装置的基本结构,阐述增殖剂换料的基本流程,对产氚辐照装置的气-固两相流进行研究,计算装置内的压力、流速、吹浮力等流场参数.结果表明,通过可靠的结构设计和合理的吹送载带气体进出口压差选择,可实现辐照装置内增殖剂的在线换料操作.     

ITER中国液态锂铅实验包层模块第一壁等离子体注入氚滞留分析

对等离子体注入ITER中国液态锂铅实验包层模块第一壁滞留的氚进行了分析,考虑了第一壁温度梯度、材料表面清洁度、加挂Be瓦及结构材料内缺陷等因素对氚滞留量的影响。分析结果显示,滞留的氚主要存在于中子辐照引起的缺陷内;氚滞留量对第一壁面向等离子体侧的清洁度及加挂Be瓦很敏感;总的氚滞留量约0.58 mg,不会对ITER真空室内氚滞留造成显著影响。     

氚/单晶硅器件辐伏电池模型及样机的长期稳定性研究

采用氚化钛源原位辐照和加速器低能电子束加速辐照实验,研究基于氚化钛/单晶硅PN结器件的氚辐伏电池模型和实验室组阵型电池原型样机的长期稳定性。测试电池模型和电池样机的氚辐伏输出随辐照时间的变化,分析辐照对单晶硅PN结型器件的本征暗特性和器件表面层材料缺陷的影响。结果表明,氚化钛源原位辐照电池模型在115 d的辐照中辐伏输出没有明显的衰减,辐照后单晶硅器件的本征暗特性曲线变化微小。电池模型的加速器低能电子束加速辐照实验表明,加速辐照在相同电子注量下对电池造成远大于氚源原位辐照的性能损伤,但损伤仅在辐照最初期快速产生,随后基本保持稳定,电子顺磁能谱（ESR）测试加速辐照60 min单晶硅器件材料的缺陷没有明显增加。组阵型实验室电池原型样机在64个月的室温储存中,基本单元的辐伏输出性能衰减比氚的自发衰变衰减有小幅增大,增大幅度小于11.4%;另外,组阵中单元之间串并联电连接有部分失效,这是后续应重点关注的问题。     

固态氚增殖剂研究进展

增殖包层作为实现可控核聚变燃料"自持"的关键,不仅能实现氚的增殖,而且起着能量转换的作用,氚增殖剂是其中最重要的功能材料。本文从材料体系的制备、性能以及改性总结了固态氚增殖剂的发展趋势。同时,基于当前的研究现状对固态氚增殖剂的发展进行了展望。     

中国氦冷固态增殖剂实验包层模块材料研究进展

在未来核聚变反应堆中,为补充氚的消耗,需要在核聚变堆的包层中进行氚的在线增殖,以维持核聚变反应的持续进行。为验证这一关键技术,在国际热核聚变实验堆（ITER）上开展了ITER TBM计划（实验包层项目）。作为ITER计划成员方之一,中方以中国氦冷固态增殖剂实验包层模块（HCCB TBM）概念参与ITER TBM计划。HCCB TBM现今进入初步设计阶段,而材料的制备技术和性能数据是支撑其结构设计、安全分析和服役工况评估的基础。本文综述和分析了HCCB TBM结构材料低活化铁素体/马氏体钢（RAFM钢）与功能材料氚增殖剂和中子倍增剂的研究现状,并对这些材料下一步的研究方向进行了展望。     

Effect of sweep gas species on tritium release behavior from lithium titanate packed bed during 14 MeV neutron irradiation

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li₂TiO₃) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.     

Impact of Resonance Treatment on Minor Actinide Incineration in a D–T Thorium Fusion Concept

Resonance treatments have an essential role to reliable neutronic calculations with different neutronic parameters. In this study presents the effect of resonance treatment and various tritium breeder materials on the incineration of the nitride fuels containing minor actinide mixed thoria in the Deuterium–Tritium fusion driven hybrid reactor as time dependent. Neutron transport calculations under resonance treatment and without resonance treatment are performed by using XSDRNPM/SCALE 5 codes. The impact of resonance treatments and various tritium breeder materials on tritium breeding, energy multiplication, total fission rate (∑_f), cumulative fissile fuel enrichment, fissile fuel breeding, average burn up values are comparatively investigated. It is observed that the neutronic results affect from both resonance treatment and the tritium breeder materials as time dependent.     

Ion beam induced luminescence of polyethylene terephthalate foils under MeV H and He ion bombardment

The evolution of the ion beam induced luminescence (IBIL) of the polyethylene terephthalate (PET) foils was studied under the irradiation of H and He ions of MeV energy. The optical and chemical changes of the samples were also examined by photo-stimulated luminescence and optical absorption measurements after the irradiation. A prominent broad emission peak of IBIL appeared at around 380 nm, and its intensity monotonically decreased during the ion irradiation. The decay curves of the emission intensity were quantitatively explained as a function of the electronic energy deposition of the incident H and He ions. On the contrary, to the decrease of the main emission peak, a growth of new peaks was observed in the wavelengths between 500 and 600 nm.     

The production of tritium in fission and fusion reactors

Conclusions It follows from the published data on the selection of lithium materials for the breeding zone of a thermonuclear reactor [43–45] that the tritium production involves large-scale radiochemical installations for the conversion of the irradiated material and the extraction of the pure isotope which is present in extreme dilution. When such radiochemical installations are built, a set of basic technological problems must be solved (continuously maintaining the purity of the lithium materials, removing the radiolysis products, stabilization of the physicochemical properties of the material, removing the corrosion products of the materials used for the construction); one must also solve problems related to personnel safety and the protection of the environment from the emission of the radioactive isotope. Problems related to the construction of apparatus for the processes occurring at a high temperature must be solved.When from this viewpoint lithium materials and their irradiation conditions are selected, one must recall that solid lithium materials (ceramics, alloys) and melts of salts have advantages over metallic lithium for the very reason that they eliminate operations involving a flammable material.The use of solid lithium materials or of molten systems as a neutron-absorbing material in the breeding zone of the blanket depends upon the characteristics of the tritium-breeding zone in terms of neutron physics and upon the thermophysical parameters.At the present time nuclear physics research, physicochemical research, and radiochemical research is done on systems with irradiated lithium materials. Methods are developed for separating the tritium which is present in various states at relatively low concentrations. The goal of this work is not only to obtain answers to the technological problems with the greatest reliability but also to effectively select a lithium material suitable for building the tritium cycle of a thermonuclear station as a whole.In the ensuing stage of the research work, when the blanket zone will be drafted, one must evaluate the various conditions of operation of the breeding zone with metallic lithium, diluted fluorine systems, and solid ceramic materials. At the present time, the selection of the blanket design and the construction of the breeding zone, as well as the selection of the lithium material providing the highest tritium production coefficient, are the basic problems among the engineering problems of thermonuclear fusion and the technological and economical aspects of the forthcoming developments.Translated from Atomnaya Énergiya, Vol. 44, No. 5, pp. 440–446, May, 1978.     

Ion beam induced luminescence on white inorganic pigments for paintings

Ion beam induced luminescence (IBIL) has been used for studying the emission features and the radiation hardness of white pigments. In particular, ZnO, gypsum and basic lead sulphate pigments have been analyzed with a 3.0 MeV H⁺ beam at the AGLAE Louvre laboratory. The same pigments mixed with different binders have been also analyzed on a canvas, in order to evaluate the contribution of the binders both to the IBIL spectra and to the radiation hardness. It turns out that the binder affects both the IBIL spectra and the radiation hardness of pigments when the emission bands are related to point defects, as occurs for ZnO.     

在线同位素分离器靶-源中靶材料的放射性活度

用LCS+CBURN程序计算在线同位素分离器靶-源中铅、钨、铜、铝、石墨靶材料以及结构材料水、不锈钢在100MeV、200μA强流质子束照射下所产生的放射性核素活度以及γ射线强度随时间的变化,以便为靶的设计、更换以及后期处理提供一定的设计依据。所选靶材料在照射后会产生长寿命放射性核素氚,其中,铅靶材料中还会产生¹³¹I。     

Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

Selection of lithium containing materials is very important in the design of a deuterium–tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid–solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.     

Tritium permeation experiment at IFMIF Medium Flux Test Module

Tritium release experiments using different breeding material candidates are planned for the medium flux region of the IFMIF Test Cell. Nowadays, only ceramic breeder materials have been suggested to be tested in the Tritium Release Module located in the Medium Flux Test Module of IFMIF. Liquid breeder blankets are very promising and for that reason, several concepts will be tested in ITER. One of the main problems concerning the liquid blankets is the permeation of the generated tritium in the breeder throughout the walls. Since tritium permeation is highly influenced by irradiation conditions, IFMIF is a suitable scenario to perform tritium permeation related experiments.In this paper, a preliminary design of a tritium permeation experiment for the Medium Flux Test Module of IFMIF is proposed, in order to contribute to the progress of the liquid breeder blanket concept validation.The conceptual design of the capsule in which the experiment will be performed is carried out, taking into consideration the experiment necessities and its implementation in the Tritium Release Module. In addition to this, some thermal hydraulic calculations have been performed to evaluate the thermal behaviour of the irradiation capsule.