反应堆退役石墨中14C分析制样实验系统研制

为分析反应堆退役废物石墨中的14C含量,设计制作了一套14C高温催化氧化制样实验系统,在实验室中对该系统的处理能力和运行功能进行了部份实验验证。结果表明：在标气流速为1L/min、催化氧化炉800℃时,对CO催化氧化能力为96%;2mol/L的NaOH溶液对CO2的吸收能力可达99%（其中,一级吸收为67%,二级32%）;空气流速为1L/min、高温解吸室850℃,1h后石墨样品分解率为99.99%;使用CaCO3固体粉末悬浮液闪测量法时,CaCO3的化学回收率为99%。     

HTR-10辐照后石墨球~(14)C含量研究

高温气冷堆中球形燃料元件与不含燃料的石墨球在14 C产生机制上基本相同,为获得10MW高温气冷堆(HTR-10)中燃料元件及石墨球中14 C的含量,研究了14 C在石墨球中的产生机理。总结了石墨球及燃料元件中14 C的产生途径,计算经过堆芯辐照后的石墨球中14 C总量,比较现有石墨球的解体技术,提出了分解石墨球制取14 C样品的实验测量方法。本文工作为进一步的实验研究工作奠定基础并提供理论计算结果比对。     

石墨中~(36)Cl分析的溶样技术

在反应堆退役过程中,为了使放射性废物最小化,需对石墨中的~(36)Cl进行分析。采用浓硫酸、浓硝酸和高锰酸钾的混合溶液溶解石墨样品,建立了石墨的溶样方法。该溶样方法对~(36)Cl的化学回收率大于89.38%。     

~(233)U内标α谱法测定石墨中痕量~(239)Pu

在石墨粉末试样中加入2 33U作内标 ,用硝酸浸取样品中的Pu ,移取上层清液制成α源。用α谱仪测定2 33U和2 39Pu的α计数比 ,通过加入内标2 33U的已知量和2 33U、2 39Pu的相关参数 ,可求得2 39Pu的绝对量。该分析方法可测2 39Pu的含量为 0 1～ 10 μg/ g ,测量精密度 (n =6)优于 2 %。     

空气中~(14)C取样器

本文介绍了空气中14 C取样器的结构、性能 ,及取样器在中辐院使用运行一年的取样监测情况。该仪器可以在高达 6～ 8L/min的空气流量下工作 ,对空气中 CO2 的收集效率≥ 98%。对于 1 0 0 min计数时间 ,空气中14 C的探测下限为 0 .0 36Bq/m3     

加速器质谱~(14)C样品制备中Zn-TiH_2法石墨产率探讨

用于加速器质谱仪(accelerator mass spectrometry,AMS)测定14 C年龄的石墨靶制备方法有很多种,提高石墨产率对优化石墨靶制备流程和改善石墨靶性能具有重要意义。本文以Zn-TiH2法制备石墨靶为例,结合相关研究,证明石墨产率与同位素分馏之间的关系;对比石墨产率的不同测算方法并提出改进建议;对实验过程中影响石墨产率的因素进行探讨,对Zn-TiH2法制备石墨靶的最优实验条件进行简要总结。     

果胶和右旋糖酐的~(14)C标记

多糖的放射性标记物是进行示踪代谢和化学结构研究等重要的工具之一。迄今为止,用~(14)CO_2的生物学方法或植物合成法较为复杂;而利用~(14)C-氰化物与多糖起缩合反应,制得~(14)C-标记的多糖,可获得满意的结果,也是目前较为实用的有效途径。     

空气中~(14)C的快速测定

本文从经济、实用的角度出发,介绍了一种用液闪技术对空气中~(14)C采取固体悬浮测量的方法。文中给出了“甲苯-Triton X-100-水”混合体系作为固体悬浮体的应用条件、计数样品及液闪测量样品的制备方法和温度影响等因素,并给出了初步的测定结果。     

固态钍基熔盐堆中~(14)C的产生及释放探讨

熔盐堆作为第四代反应堆论坛推荐的6种候选堆型之一,具有输出温度高、能量密度高、无水冷却等特点。固态钍基熔盐堆(Thorium Molten Salt Reactor with Solid Fuel,TMSR-SF1)堆芯大部分结构材料为石墨,冷却剂杂质及石墨材料中的13C和杂质N、O易被活化产生14C。14C半衰期较长,同其他稳态核素12C、13C一样广泛参与各种复杂的生物循环,在反应堆中受到关注。TMSR-SF1中的14C广泛分布于冷却剂、堆芯石墨结构材料和燃料元件。本文采用输运燃耗耦合方法,应用SCALE6.1的TRITION控制模块对反应堆各区域的14C放射性活度进行计算分析,结果表明,反应堆在正常运行工况下一回路每年产生的14C放射性活度为0.34 TBq,满足现有的压水堆、重水堆管理限值要求。向环境释放的14C主要来自于一回路熔盐中N杂质的活化。     

大气和生物样品中~(14)C分析方法的研究

本文叙述一种监测大气和生物样品中~(14)C的分析方法,包括由样品制备CO_2,CO_2经Li_2C_2制成C_2H_2,C_2H_2被VrO_3活化的Si-Al催化剂转化为苯。苯的产额大于80％,经气相色谱分析鉴定苯的纯度优于色谱级苯。方法简便、快速、准确,适用于监测环境样品中~(14)C。     

基于SAXS方法的核石墨微孔结构的温度影响

X射线小角散射(Small Angle X-ray Scattering,SAXS)是研究纳米尺度微观结构的重要手段。本文利用同步辐射SAXS技术测量了25oC、100oC、200oC、300oC和400oC时,IG-110和NBG-18核石墨在纳米尺度范围内孔隙的数量分布及其分形特征的变化。实验结果表明,IG-110和NBG-18核石墨的微观结构中存在微小尺寸上的不均匀区域,且核石墨孔隙的固气结构具有明锐的界面。但随着温度的升高,固气界面的变化并没有呈现出明显的规律性。此外,在纳米尺度上,IG-110和NBG-18核石墨的孔隙数量随温度呈现增加的趋势,且IG-110核石墨孔隙数量的增加幅度大于NBG-18核石墨,其平均孔隙尺寸的减小幅度大于NBG-18核石墨。在核石墨的微孔结构内,其固气界面的分形维数随温度升高逐渐减小,且NBG-18核石墨分形维数的变化幅度小于IG-110核石墨。这表明核石墨的分形结构随温度的升高逐渐光滑。     

Experimental study on the adsorption and desorption of tritium in the graphite materials for HTR-PM

Tritium behavior in the reactor such as production, diffusion and release are accompanied by their adsorption and desorption in graphite materials, which are essential to the safety of high temperature gas cooled reactor (HTGR). In order to study this important issue, hydrogen instead of tritium is experimentally used in this work and justified viable by theory. By performing multiple sets of comparative experiments, the features of hydrogen adsorption and desorption behavior changing by adsorption temperature and time in typical graphites used in HTR-PM (High Temperature Gas Cooled Reactor – Pebble Bed Module), i.e. reflective layer, fuel element and boron carbon bricks, have been observed and analyzed. Furthermore, the adsorption rates of hydrogen in the three materials as above at different conditions are also given. Based on the experimental results, tritium behavior in the HTR-PM was inferred and estimated, which is significant for the further study on the mechanism of tritium transport.     

Quantitative determination of the amount of3H and14C in reactor graphite

Moscow Engineering Physics Institute (MIFI). Kurchatov Institute (RNTs). Translated from Atomnaya Énergiya, Vol. 73, No. 6, pp. 446-450, December, 1992.     

Adsorption of iodine on graphite in High Temperature Gas-Cooled Reactor systems: A review

Adsorption of iodine on graphite is of great interest for operation and safety of high temperature nuclear reactors. Graphite can adsorb significant amounts of iodine and retain it for a long period of time. Significant amount of work on this subject has been done in the past. Various types of adsorption apparatus have been designed and data were collected. The types of graphite used in past studies are not available anymore, and as a consequence the data are not applicable for the new type of commercial nuclear grade graphites. However, the past experimental systems, data, and their analysis are useful to design a better experimental system, collect more accurate data, and, provide better understanding of the adsorption process and data. In addition the existing data can be used to generate a framework to understand the types of adsorption processes taking place. In this work, we have conducted an exhaustive literature review and further analyzed the data. Four adsorption isotherms; the Langmuir, the Freundlich, and the two isotherms proposed in the International Atomic Energy Agency (IAEA) Tecdoc-978 were used to correlate the available equilibrium adsorption data. For most of the data, the simple Langmuir and the Freundlich isotherms provided a reasonable fit of the data. The Polyani's potential theory was also used to check the consistency of the data and as indicated by the theory, most of the data set provided a single characteristic curve. The isosteric heats of adsorption calculated using the literature data suggested that iodine-adsorption on graphite could be a chemisorption process.     

Velocity effect of swift heavy ions in graphite studied by Raman spectroscopy

Highly oriented pyrolytic graphite (HOPG) samples were irradiated with swift heavy ions (Ar, Kr, Bi, U) of fluences between 10¹¹ and 10¹³ ions/cm² in energy range MeV-GeV. The irradiated samples were analyzed by Raman spectroscopy with laser wavelength of 532.2 nm. It is shown that the ratio between the integrated intensities of the disorder-induced D and the original G Raman bands which denotes the degree of the damage induced by ion irradiation increases as a function of ion fluence as well as the electronic energy loss. This agrees with the previous reports. However, quantitative analysis of the peak intensity at a fixed fluence discloses that ion velocity is also a significant parameter in determination of damage. The conclusion is that the extent of discontinuity of ion track may change with ion velocity besides the electronic energy loss. Considering the radial distribution of the energy deposited on the matter being velocity dependent, the energy density which combines the influence of the electronic energy loss and ion velocity may be more suitable for explaining the effect induced by swift heavy ions.     

Saturation in deuterium retention of CFC graphite targets exposed to beryllium-seeded plasmas on PISCES-B

ITER strike-plates are foreseen to be of carbon-fiber-composite (CFC). In this study the CFC bulk deuterium retention in ITER-relevant conditions is investigated. DMS 701 (Dunlop) CFC targets were exposed to plasma in PISCES-B divertor plasma simulator. Samples were exposed to both pure deuterium plasma and beryllium-seeded plasma at high fluences (up to ) and high surface temperature (1070 K). The deuterium contents of the exposed samples have been measured using both thermal-desorption-spectrometry (TDS) during baking at 1400 K and ion beam nuclear reaction analysis (NRA). The total deuterium inventory has been obtained from TDS while NRA measured the deuterium depth distribution. In the analysed fluence range at target temperature of 1070 K, no fluence dependence was observed. The measured released deuterium is . In the case of target exposure with beryllium-seeded plasma no change in the released amount of deuterium was found. The deuterium concentration inside the samples is almost constant until the probed depth of ?m, except in the first 1 μm surface layer, where it is 5 times higher than in the bulk. No C erosion/redeposition was observed in the Be-seeded plasma cases. The measured retention, applied to 50 m² of ITER CFC surface, would imply a tritium saturated value of 0.3 gT, much lower than the ITER safety limit of 350 g.     

Radiation and thermal effects on cobalt retention by Mexican aluminosilicates

Thermal and radiation effects on the leaching of cobalt from two cobalt exchanged zeolites and one clay were determined. The cobalt exchanged aluminosilicates were heated at different temperatures (500, 700, 900 and 1100 °C), and the materials were then treated with NaCl (1 and 5 M) and HNO₃ (0.001 and 1 M) solutions to determine the leaching behavior of cobalt from the materials. Cobalt showed greater stability when the materials were heated at the highest temperature. The unheated samples and those heated at 1100 °C were gamma irradiated, and it was found that cobalt leaching from gamma irradiated aluminosilicates was higher than that for non-irradiated materials.     

基于压水堆运行反馈的~(14)C源项研究

本文阐述了压水堆中~(14)C产生机理,建立了~(14)C产生量的计算模型和方法。通过对德国和法国大量压水堆的气相~(14)C排放量进行统计分析,得到法国和德国压水堆的气相~(14)C年排放量平均值为217GBq/(GWe·a),提出气相~(14)C最大排放量可取平均值的1.4~1.7倍的经验方法。结合理论计算,指出固相和液相~(14)C可能占~(14)C总产生量的20%以上。研究表明,引起同类压水堆中气相~(14) C年排放量在较大范围变化的主要原因是机组运行中放射性废气排放管理的不确定性,而不是由于冷却剂氮浓度变化。本文的研究方法和结论对于压水堆设计具有普遍适用性,可用于三代压水堆的放射性流出物设计和工程评审。