大面积氘/氚靶制备及中子产额分析

大面积氘/氚靶是实现高产额强流中子源的关键部件,是氘、氚中子源广泛应用的前提条件。本工作采用磁控溅射镀膜与多弧离子镀结合的方式,制备以铜或钼为基底、直径大于500 mm的大面积钛膜。针对制备的钛膜,采用自研的氘/氚靶生产系统,经过除气、净化、高温吸氘/氚、尾气回收等流程,生产了氘/氚钛原子比大于1.85的氘靶、氚靶,采用Ф22 mm的小靶片,进行氘束流加速器中子产额测试,研制的氘靶中子产额达到8.0×10⁸/s,根据氘靶与氚靶反应截面计算氚靶中子产额,相同条件下,氚靶的中子产额在1.0×10¹¹/s以上。以上测试结果表明,本工作研制的Ф500 mm大面积氘/氚靶,可实现强流中子源的高产额中子输出,达到国际先进水平。     

加速器含氚废气净化系统的技术指标估算

根据加速器应用环境的要求,对建立纳秒脉冲/直流强中子发生器的含氚废气净化处理系统的主要参数、性能和技术指标进行了较为详细的分析和估算。含氚废气排放量大于4 m3/d,最大含氚浓度约1×1012Bq/m3,净化系统的除氚因子应大于1×103。建议采用国内外广泛使用的催化氧化加分子筛吸附的净化技术;结构上采用进出口都有氚浓度在线监测的装置,全自动智能控制加手动控制的三级串联的净化系统。     

可移动式除氚器的现场应用试验

在现场闭路循环工作模式下，采用中国辐射防护研究院研制的小型可移动式除氚器对微量(（0.5~1）×10^-6)气态氚进行脱氚试验。试验结果表明：催化床在450℃工作温度下，空气闭路循环流量为3.8m³/h时，除氚器可在114min内将1m³密封不锈钢罐空气中的氚浓度由4.62×10⁶Bq/L降至4.62×10³Bq/L。同时，论证了除氚器在现场使用的适应性与安全性，并对其研制和现场性能试验中发现的技术问题进行了论述。     

氚纯化装置的研制及性能验证研究

铀是一种传统的贮氚材料,在铀粉瓶中贮存的氚会不断衰变产生氦气,导致使用时氚的纯度下降,影响标记化合物产率。本研究设计了氚纯化装置,对装置进行安装调试,并对该装置中的铀床进行活化,利用该装置测定铀吸收氘单质气体的p-t曲线及在400~550℃范围的解吸氘气体的p-t曲线。应用调试好的系统对长期存放贮氚铀粉瓶中的氚进行纯化。结果表明,设计的氚纯化装置系统密封性好,经氦质谱检漏测定值为7.8×10~(-13 )Pa·m~3/s;利用该装置测定氘的吸附饱和曲线,氘完全解吸时铀对氘的吸附量为240mL/g。验证实验回收了久置铀粉瓶中的氚为1.44×10~(13 )Bq,利用氦气体积推算出久置铀粉瓶中含氚质量百分率为53.1%。实验结果证实了系统纯化氚的可行性,可为氚标记化合物制备提供可靠的氚源。     

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。     

强流氘氚聚变中子源HINEG设计研究

强流氘氚聚变中子源HINEG(High Intensity D-T Fusion Neutron Generator)研发分两期:HINEG-Ⅰ为直流脉冲双模式,已成功产生中子强度1.1×10~(12)n/s的氘氚聚变中子,并实现连续稳定运行;HINEG-Ⅱ中子强度设计指标为10~(14)~10~(15)n/s量级,重点突破强流离子源和高载热氚靶技术。HNEG中子源可开展中子学方法程序与核数据、辐射屏蔽与防护、材料活化与辐照损伤机理和部件中子学性能等核能与核安全研究,同时也可在核医学与放射治疗、中子照相等领域拓展核技术应用研究。本文简要介绍HINEG总体设计方案与关键技术研究进展。     

TBM TES手套箱泄漏事故下氚浓度与通风关系研究

为了分析TBM TES室内通风是否能在手套箱发生氚泄漏时控制室内氚浓度在安全剂量值,采用Fluent对手套箱的氚泄漏和扩散进行模拟,得到不同通风下手套箱泄漏时的氚泄漏速率以及室内氚浓度分布,并对比分析了模拟数据与理论数据。结果表明TES手套箱泄漏速率为1.41×10~(-5)g/s和1.02×10~(-5)g/s时,分别为5次/h和8次/h的换气通风能控制室内氚浓度在安全剂量值2.0×10~(10)Bq/m~3内;而氚泄漏速率为1.56×10~(-5)g/s时,3次/h的换气通风不能控制室内氚浓度在安全剂量值内;模拟结果与理论结果相一致。结果为TES通风除氚设计提供了理论依据。     

利用伴随α粒子能谱分析固体氚靶中氚浓度深度分布

在目前的氘氚中子发生器源中子分析过程中,固体氚靶中氚浓度深度分布信息的缺失是普遍遇到的问题。为解决此问题,本文建立了利用伴随粒子能谱反演氚浓度深度分布的模型,采用来自氚钛靶的α实验能谱作为模型测试对象,通过该模型获得了氚钛靶中氚浓度深度分布的数据。结果表明,氚浓度随氚钛靶深度的增加呈双峰趋势,两峰之间的氚浓度波谷位于靶中0.94 μm处,该深度正是入射氘粒子的射程极限。所得的氚浓度深度分布趋势与其他实验方法测量结果相符,表明该模型能为氘氚中子发生器的源中子分析提供即时的氚浓度深度分布信息。     

氚纯化装置的研制及性能验证研究

铀是一种传统的贮氚材料,在铀粉瓶中贮存的氚会不断衰变产生氦气,导致使用时氚的纯度下降,影响标记化合物产率。本研究设计了氚纯化装置,对装置进行安装调试,并对该装置中的铀床进行活化,利用该装置测定铀吸收氘单质气体的p-t曲线及在400~550 ℃范围的解吸氘气体的p-t曲线。应用调试好的系统对长期存放贮氚铀粉瓶中的氚进行纯化。结果表明,设计的氚纯化装置系统密封性好,经氦质谱检漏测定值为7.8×10^-13 Pa•m³/s;利用该装置测定氘的吸附饱和曲线,氘完全解吸时铀对氘的吸附量为240 mL/g。验证实验回收了久置铀粉瓶中的氚为1.44×10¹³ Bq,利用氦气体积推算出久置铀粉瓶中含氚质量百分率为53.1%。实验结果证实了系统纯化氚的可行性,可为氚标记化合物制备提供可靠的氚源。     

气相色谱法浓缩氘的研究

采用气相色谱技术建立了一套含氚重水氘浓缩的演示实验装置,用氘-氢（D-H）体系进行实验,结果表明,装置实现了长期、连续、安全的运行。D浓度为O．1％的样品经装置浓缩后得到D浓度为30％的产品,贫化部分D浓度小于10^-5;D浓度为10％的样品经装置浓缩后得到D浓度大于90％的产品。装置对D-H体系的处理容量达到40m%3/d。     

Preliminary results from a detritiation facility dedicated to soft housekeeping waste and tritium valorization

Nuclear waste management has to be taken into account for fusion machine using tritium as fuel. Soft housekeeping waste (e.g. gloves, tissues, protective clothes, etc.) is produced during the whole life as well as during the dismantling of the reactor and is contaminated by tritium under reduced (HT) and oxidized (HTO) forms.In collaboration with ENEA, a lab-scaled facility has been built at CEA Cadarache for soft housekeeping waste detritiation and tritium valorization. The previously milled waste is placed in a reactor to be heated up to a temperature lower than the housekeeping melting point. A carrier gas is then injected in the detritiation reactor to remove tritium, thanks to the combined effects of temperature and carrier gas (type and feed flow). The tritiated gas exhausted from the detritiation reactor is then sent through a catalytic Pd–Ag membrane reactor (CMR) where tritium is recovered via isotopic exchange reaction and permeation phenomenon.Based on previous studies that have allowed defining the most efficient operating conditions for the detritiation process, this work presents the results obtained by the coupling of the detritiation facility with the CMR. Due to safety considerations, restrictions on the nature of the carrier gas were applied, rejecting air as the carrier gas even though air was the best candidate for the detritiation part of the process. The performance of the whole system was estimated by means of a parametric study on the influence of flow rates in the CMR and transmembrane pressure.     

Effect of hydrocarbons on the efficiency of catalytic reactor of detritiation system in an event of fire

Detritiation system of a nuclear fusion plant is mandatory to be designed and qualified taking carefully into consideration all the possible extraordinary situations in addition to that in a normal condition. We focused on the change in the efficiency of tritium oxidation of a catalytic reactor in an event of fire where the air accompanied with hydrocarbons, water vapor, and tritium is fed into a catalytic reactor at the same time. Our test results on the effect of these gases on the efficiency of tritium oxidation of the catalytic reactor indicated; (1) tritiated hydrocarbon produces significantly by reaction between tritium and hydrocarbons in a catalytic reactor; (2) there is little possibility of degradation in the detritiation performance because the tritiated hydrocarbons produced in the catalyst reactor are combusted; (3) there is no possibility of uncontrollable rise in the temperature of the catalytic reactor by heat of reactions; and (4) saturated water vapor could temporarily poison the catalyst and degrades the detritiation performance. Our investigation indicated a saturated water vapor condition without hydrocarbons would be the dominant scenario to determine the amount of catalyst for the design of catalytic reactor of the detritiation system.     

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.     

联合电解催化交换系统的动态模型及理论计算

为探求联合电解催化交换系统各单元中氚浓度空间分布和动态变化的内在规律，建立了D／T体系的气-液两元模型。根据不同的催化剂传质性能，计算了为达到特定脱氚率和电解池浓缩倍数所要求的交换床总高度和进液位置。理论计算得到的氚在交换床上的空间分布趋势与文献报道的中试结果一致，电解池中的氚浓度随时间呈线性增长。     

Tritium Safety-Related Studies at TPL of JAERI

Activities regarding tritium safety technology in the Tritium Process Laboratory (TPL) at Tokai Establishment of Japan Atomic Energy Research Institute are reviewed. Research and development of a new tritium removal system is being carried out by using a gas separation membrane which enable to make the ITER atmosphere detritiation system more compact and cost-effective. Techniques of gas flowing calorimetry and laser Raman spectroscopy are applied to develop new tritium accountancy methods. Studies of tritium-material interaction, such as plasma material interactions, radiochemical reaction of tritium in gas phase, radiolysis of tritiated water, and waste processing are being carried out under ITER/EDA and U.S.-Japan collaboration. Tritium release experiments for research of tritium behavior in confinements and environment and demonstration of safety related components are planned.     

Influence on moisture and hydrocarbons on conversion rate of tritium in catalytic reactors of fusion-DEMO detritiation system

Thoughtful consideration of abnormal events such as fire is required to design and qualify a detritiation system (DS) of a nuclear fusion facility. Since conversion of tritium to tritiated vapor over catalyst is the key process of the DS, it is indispensable to evaluate the effect of excess moisture and hydrocarbons produced by combustion of cables on tritium conversion rate considering fire events. We conducted demonstration tests on tritium conversion under the following representative conditions: (I) leakage of tritium, (II) leakage of tritium plus moisture, and (III) leakage of tritium plus hydrocarbons. Detritiation behavior in the simulated room was assessed, and the amount of catalyst to fulfill the requirement on tritium conversion rate was evaluated. The dominant parameters for detritiation are the concentration of hydrogen in air and catalyst temperature. The tritium in the simulated room was decreased for condition (I) following ventilation theory. An initial reduction in conversion rate was measured for condition (II). To recover the reduction smoothly, it is suggested to optimize the power of preheater. An increase in catalyst temperature by heat of reaction of hydrocarbon combustion was evaluated for condition (III). The heat balance of catalytic reactor is a point to be carefully investigated to avoid runaway of catalyst temperature.     

Design Study of Feasible Water Detritiation Systems for Fusion Reactor of ITER Scale

Two types of water detritiation systems have been designed for fusion reactors of ITER scale. One of the systems is a combination of WD (Water Distillation) and VPCE(vapor phase catalytic exchange) columns. The other is a combination of a WD column and a CECE(combined electrolysis catalytic exchange) column. Three water distillation columns are needed for the former system. The total height of the three columns is 106 m. The height of the water distillation and CECE columns for the latter system are 20 and 24m, respectively. These large water distillation columns result in the larger tritium inventory of the former system than for the latter system. However, there have been the results for the operation of the actual scale of the water distillation and VPCE columns. No demonstration test has been carried out for the CECE column. From these reasons, the WD+VPCE system should be the first candidate for the fusion reactor. The WD+CECE system is superior to the WD+VPCE system for the flexibility in design as well as the tritium inventory. It is desired to demonstrate the CECE column to develop the water detritiation system best suited to the fusion reactors.     

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.     

Preliminary results from a detritiation facility dedicated to soft housekeeping waste

Nuclear waste management has to be taken into account for fusion machine in tritium experimentations. Soft housekeeping waste is produced during both operating and dismantling phases and is contaminated by tritium under reduced (HT) and oxidized (HTO) forms. At CEA Cadarache, a lab-scaled facility has been built for soft housekeeping detritiation. The tritiated gas exhausted from the process described above is foreseen to be treated by a tubular Pd–Ag membrane reactor, for gaseous tritium recovery. Since this membrane reactor uses hydrogen as swamping gas the compatibility toward explosive hazard has to be taken into account. Then, this work presents a double objective. A first study is presented in order to identify the best conditions for the declassification of soft housekeeping waste, without tritium recovery. Experiments carried out at 120 °C are not efficient enough and do not allow one to choose the most efficient carrier gas. Some other tests are being currently performed at higher temperatures (150 °C). Moreover, due to safety issues, the use of air has to be avoided during membrane reactor implementation phase. Preliminary results obtained with hydrogen hazard-free carrier gases are also presented.     

Detritiation of Glovebox Atmosphere by Using Compact Tritium Removal Equipment

A compact tritium removal equipment (TRE), assembled in a console with casters, has been developed for detritiation of air in a glovebox used for handling of several curies of tritium. The TRE was designed to remove gaseous tritium in the form of T₂, HT and CH₃T through oxidation with precious metal/alumina catalysts followed by adsorption on zeolite pellets.

From the detritiation experiments with hydrogen tritide (HT, 2–20 mCi), the TRE was confirmed to have sufficient performance for the practical use. The tritium concentration in the test gas (total volume –32l; 1%H₂, 5%O₂, 94%N₂) decreased from 0.64 to 6.4 ×10-⁷ Ci.m³ within 155 min when the TRE was operated under the recirculation mode with the flow rate of 200 l-h¹ at the catalyst temperature of 200°C. In addition, the HT-to-HTO fractional conversion was determined at various catalyst temperatures (25–200°C) and flow rates (100–360 lh-¹).