核分析技术研究一种水系沉积物的多元素取样行为

为给研制适用于固体取样微分析质控的天然基体标准物质提供可能的途径，利用3种核分析技术（INAA、PIXE和SRXRF）对自行研制的一种水系沉积物在取样量跨越9个量级（10^-1～10^-10 g）水平进行多元素取样行为研究。实验结果表明，该基体在不同取样量下，一批元素具有满意的取样不确定度。     

Electrochemical oxygen sensors for on-line monitoring in lead-bismuth alloys: status of development

This paper presents the state of development of oxygen sensors based on the electromotive force (emf) measurement at null current, using yttria stabilized zirconia as solid electrolyte for application in liquid lead-bismuth eutectic (LBE), which is envisaged as a nuclear coolant or as a spallation target in accelerator driven system (ADS) for nuclear waste transmutation. The assembly procedure, the calibration method, as well as the summary of the various validation tests undergone in both static and loop facilities are presented so as to define a real state of achievement and the basics needs for further studies. The sensors are efficient, accurate, rapid and reliable for research loops. However, the poor mechanical resistance as well as the effect of traces of impurities, promoting an increasing time-drift under certain conditions, are to be further studied to improve the sensor reliability for a nuclear use. The oxygen and chromium solubilities were reassessed in the process of the sensor testing, those relations are also given and discussed.     

Removal of Liquid and Solid Impurities from Uranyl Nitrate Hexahydrate Crystalline Particles in Crystal Purification Process

The purification behavior of uranyl nitrate hexahydrate (UNH) was investigated to evaluate the decontamination performance of liquid and solid impurities using a dissolver solution of mixed oxide (MOX) fuel in batch experiments. The UNH crystal recovered from the MOX fuel dissolver solution containing simulated fission products (FPs) was purified by a sweating and melt filtration process. Although the decontamination factors (DFs) of Pu, Cs, and Ba did not change in the sweating process, that of Eu increased with increases in temperature and time. These results indicate that liquid impurities such as Eu were effectively removed by the sweating method, but solid impurities such as Pu, Cs, and Ba were minimally affected in the batch experiments. On the other hand, the DF of Ba increased with 0.45 and 5.0 μ filters in the melt filtration process. Since Pu and Cs formed as Cs₂Pu(NO₃)₆ in the course of U crystallization and was accompanied with the UNH crystal, these behaviors were similar to each other. Although the DFs of Pu and Cs did not change with the 5.0 μ filter, it increased approximately twofold with the 0.45 μ filter. The particle size of Cs₂Pu(NO₃)₆ is relatively small and might pass through the 5.0 μ filter in the melt filtration process. The liquid impurities as Eu remained in the molten UNH crystal with some filters.     

Gas/liquid oxygen-transfer to flowing lead alloys

For the application of liquid lead or lead-bismuth eutectic as coolant in nuclear reactors, the concentration of dissolved oxygen, determining the compatibility with steels used as construction materials, is of critical importance. In general, oxygen has to be added continuously to the liquid metal, so as to compensate for consumption by oxide formation on the surface of the reactor components. A potential method of keeping the oxygen concentration in a favourable range is transferring oxygen from an oxygen-containing gas, which is investigated on the basis of the experience from operating a gas/liquid transfer device in the CORRIDA loop. Conclusions on oxygen transfer in industrial-scale systems are drawn.     

Electrochemical techniques for monitoring dissolved carbon, hydrogen and oxygen in liquid sodium

Potential chemical problems associated with the use of liquid sodium as a coolant for fast reactors have led to the need for special analytical techniques, in particular, for dissolved carbon, hydrogen and oxygen. Consideration of the chemical monitoring requirements involved shows methods based on electrochemical cells to be particularly suitable. After a summary of the relevant background phenomena, the development and operational experience of specific electrochemical carbon, hydrogen and oxygen meters are described.     

液态铅铋合金回路氧输运特性的数值研究

为研究液态铅铋合金（LBE）冷却剂系统气态氧控装置——膨胀箱中覆盖气体的氧输运特性,利用计算流体动力学（CFD）软件ANSYS Fluent对氧输运进行了数值计算。根据覆盖气体流动特性和混合气体中低氧分压特点,对膨胀箱气相空间进行简化,将气-液交界面视为氧浓度恒定的自由表面边界,采用组分输运模型计算气体和液态LBE之间传质后的液态LBE氧浓度。结果表明,传质系数随液态LBE入口流速增大而增大,液态LBE入口流速增大则膨胀箱内气-液对流强度增加,有利于增强膨胀箱的氧输运;膨胀箱中液态LBE温度越高,则氧输运的平均传质系数越大;在液态LBE入口流速一定时,平均传质系数可表示为温度的递增函数。在饱和氧浓度阈值内,入口氧浓度和气-液交界面氧浓度不影响膨胀箱的传质系数,对液态LBE回路的氧浓度控制有利。本研究定量获得了使液态LBE回路处于合理氧浓度范围内的操作条件,为实验及系统设计提供数据参考。     

A proposal for an aseismic design method of equipment and piping for NPPs in a low seismicity area

Recently a regulatory code for an aseismic design of high-pressure gas facilities became effective by the order of the Ministry of International Trade and Industry (MITI) in Japan. This order includes details of the aseismic design of vessels whose “factor of importance” are relatively lower than Class A (Class I) items in nuclear power plants.The author develops his idea on an aseismic design method of equipment and piping of nuclear power plants in a Low Seismicity Area (LSA) based on his experience of the new code for petro-chemical industries and oil refinaries pertaining to high pressure gas facilities mentioned above.The definition of LSA is usually the area whose maximum intensity has never exceeded MMI VI or VII. However, there are two types of LSA, one is really such a low seismicity area, and the other type is the area which has the possibility of stronger earthquake occurrence than those mentioned above, even though it is low. One of the typical examples is the area subjected to “New Madrid Earthquake-1812”. The author develops his concept along these two lines.He briefly describes the new code for high-pressure gas facilities in Japan. This code describes the design methodology of both types of aseismic design analysis, that is, rather sophisticated dynamic methods for facilities whose potential hazard is as high as those in a nuclear power plant, such as liquified chlorine gas storage, and simplified dynamic and static methods for most of the equipment and vessels in those plants. One of the features of this code is the use of design formulae and charts to simplify their design procedure as well as the set of specific computer codes by the MITI. These computer codes are prepared by the MITI or approved by the MITI for providing equivalent capability to the practice designated in the MITI order.The author's philosophy for the code of equipment and pipings in LSA is that they must be as simple as possible, and most of the analytical work for the design should be eliminated, or at least limit the use of simplified methods, such as the static seismic coefficient method or the modified seismic coefficient method with a simplified response spectrum. The use of general design criteria or a guideline of structural details may be better than a sophisticated design analysis as a result.     

岩石和矿物中氧同位素组成分析制样装置改进及分析方法

对五氟化溴法氧同位素组成样品制备装置进行改进并对分析方法进行优化。改进的样品制备装置包含12个镍反应器,可完成12件氧同位素样品制备,提高了制备效率;采用O 2作为工作气体对样品中的氧同位素组成进行测定,避免了向CO 2转化过程中潜在的氧同位素分馏,还可同时获取δ18 O与δ17 O值;通过5A分子筛直接对O 2进行吸附,提高了收集效率;采用电容真空规对提取的O 2进行产率测量,确定了制样装置中石英单矿物质量与O 2压强之间的线性关系。以国家标准物质GBW 04409、石英单矿物以及花岗岩地质样品为参考,在改进的装置上进行氧同位素样品制备分析,其分析精度分别为0.09‰、0.09‰和0.10‰,均优于0.2‰,标准物质δ18 O VSMOW的测量平均值为11.09‰,在误差范围内与标准值11.11‰相吻合。样品制备效率及分析结果表明:改进的氧同位素样品制备装置及分析方法测量结果准确。     

An XPS study on the evolution of type 304 stainless steel surface during routine TCSU operation

A surface analysis study was carried out to monitor the first-wall evolution in the Translation, Confinement, and Sustainment Upgrade (TCSU) experiment. A type 304 stainless steel sample was exposed to processes including the standard ex-situ surface preparation, helium glow discharge cleaning (He-GDC), plasma discharges, and backfilling the vacuum chamber with filtered N₂. After each process, the sample was carried to a surface analysis chamber for X-ray photoelectron spectroscopy (XPS), using a custom designed in-vacuum transfer device. Results indicated that He-GDC was effective in removing both physically and chemically bound carbon and oxygen on the stainless steel surfaces due to the physical impact of the glow. The plasma discharges resulted in oxidation on the surface. The use of filtered nitrogen during vacuum breaks was verified as an effective method for minimizing carbon and oxygen contamination.     

An evaluation of the influence of impurity concentration and composition in concrete on activation and radiation dosein the reactor facility

The content analysis of radioactive waste and radiation dose evaluation is considered as one of the important factors in the reactor facility design.This kind of buildings consists of the concrete for the most part and uses it as the structure and shield of the building.Generally,the concrete has impurities such as cobalt,europium,nickel,and cesium with specific content depending on the production method or manufacturing company.Dominant radioactive nuclides generated from the fundamental components of concrete are considered that it is less contributed to the radiation dose because they are beta decay nuclides in general.Thus,impurities of irradiated concrete in the reactor facilities,are considered occasionally an important evaluation factor for induced activity.In this study,the influence on the activation of impurities in concrete was evaluated from the radiation dose and induced activity calculations.The calculation was evaluated at the bio-shield which is one of the areas with the highest neutron irradiation among the concrete structure in the reactor facility.The results show that radioactive nuclides with gamma decay were produced in these impurities.Moreover,the radiation dose of concrete with impurities was higher than concrete without impurities.The increased radiation dose was quantified through the content of impurities.     

快堆钠中杂质分析和监测技术综述

针对中国实验快堆(CEFR)冷却剂钠中杂质的质量标准,系统地概述了钠中微量氧、钙、碳、钾、氮、氯、铁杂质的离线分析方法以及阻塞计的在线监测技术。     

STIMULATING EFFECT OF LOW DOSE ~(147)Pm ON DNA REPAIR IN SPERMIOGENIC STAGES

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伺服管主导型控制棒水力驱动机构静态特性理论分析

本文对伺服管主导型控制棒水力驱动机构的静态保持特性进行理论计算和分析,得到了水力驱动缸的静态保持流量及其随温度的变化规律。结果表明：在可变节流口间隙的稳态工作范围内,系统所需的静态保持流量很小;随温度的升高,流体密度降低,静态保持流量相应增大。倾斜工况下的系统静态保持流量范围较正常工况下的大,控制棒的保持特性更加稳定,抗扰动能力更强,满足舰船核动力装置规范的规定。     

Fabrication and characterization of Li4SiO4 pebbles by melt spraying method

The solid breeder blanket concept proposed by the China features the tritium breeding ceramic as pebble beds in several submodules. The lithium orthosilicate (Li₄SiO₄) is considered as first candidate ceramic breeder materials fabricated by the melt spraying method, which is favorable to other processes in terms of density and recycling. The production process involves rapid quenching of the liquid droplets from the melt to room temperature which cause internal stresses and leads in some cases to formation of microcracks and the dispersion of mechanical properties. Molar ratio (Li/Si) of the pebbles was evaluated by ICP–OES. It is shown that the Li/Si ratio of the pebbles is slightly varying from batch to batch because of evaporation of lithium at high temperatures. The crush tests on single pebbles show that a mean value of 7.0 N was obtained in crush load experiments of 40 pebbles with a diameter of 1.0 mm. It results that heat treatment of pebbles improves the density and mechanical stability. The activation characteristics for the current composition of Li₄SiO₄ pebbles were assessed. The calculations were used to identify critical amounts of impurities and were compared to the results of pure material without impurities.     

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany.In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented.Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact.A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.     

Development of an effective dose calculation method for the downstairs room of a linac facility

ABSTRACT

An effective dose calculation method is important in the design of efficient shields in radiation facilities. Some analytical methods have been shown to provide a simple and quick design analysis; however, no suitable method exists that can be applied to a room located directly under an X-ray irradiation room. We propose a new analytical method that uses the multiple reflection ratio predetermined by a Monte Carlo simulation and the differential dose albedo given by the Chilton–Huddleston semi-empirical equation. Our method is verified by comparison with the Monte Carlo simulation, performed for the case of an electron linac facility with an accelerated energy of 10 MeV, where the shielding floor has a thickness of 1.6–2.0 m and the downstairs room has a height of 0.5–1.5 m. The difference between the effective X-ray doses in the downstairs room calculated via the proposed analytical method and the Monte Carlo simulation is less than 25% when the horizontal distance from the X-ray beam to the evaluation point exceeds 3 m and the evaluation point is set at half of the height of the room. The new analytical method can be efficiently and accurately applied to the calculation of the effective dose.     

Flow measurement techniques in heavy liquid metals

This paper summarizes several flow measurement systems qualified in the operation of different lead-bismuth loops in the KArslruhe Liquid Metal LAboratory (KALLA) during the last 5 years. There are several experimental techniques which were well proven in air and water and thus could be transferred similarly to liquid metals: these techniques are split into measuring local quantities as temperature, pressure e.g. by means of pressure taps or velocities using Pitot and Prandtl tubes or the Ultrasound Doppler velocimetry (UDV) for local flow velocities, as well as an integral quantity like the flow rate. Since the knowledge of the flow rate acts in terms of the operational safety of nuclear liquid metal systems as one of the most crucial parameters, this aspect is discussed widely herein. Unfortunately, as liquid metals are opaque, an optical access is not possible. Instead, one can take advantage of the high electric conductivity of liquid metals to measure integral and local quantities, like electromagnetic flow meters and miniaturised permanent magnetic probes for local velocity measurements. In this context especially the electromagnetic frequency flow meter (EMFM) is discussed as a prospective and reliable option to measure the flow rate without demanding extensive precognitions with respect to the fluid-wall interface behaviour.This article describes some of the techniques used in KALLA for different liquid metals, explains the measurement principle and shows some of the typical results obtained using these techniques. Also the measurement accuracy as well as the temporal and spatial resolution of each device is discussed and typical error sources to be expected are illuminated. Moreover, some hints for a correct placement of the individual sensor in the liquid metal environment are given.     

Influence of Impurities on Carrier Removal and Annealing in Neutron-Irradiated Silicon

The influence of impurities on carrier removal and annealing has been investigated in neutron-irradiated silicon in the resistivity range from 0.5 to 50 ohm-cm. Carrier removal rates in n-type material are strongly dependent upon the crystal growth method and are lower in Czochralski-grown (oxygen containing) material than in material grown by the vacuum-float-zone or LOPEX techniques. A slight dependence of the removal rate on the dopant impurity is observed in vacuum-float-zone material but not in Czochralski-grown material. The annealing behavior of n-type material is also very crystal growth dependent. An annealing stage located between approximately 144°C and 170° C is observed in vacuum-float-zone and LOPEX-grown material but not in material grown by the Czochralski method. The location of the stage is dependent upon the dopant impurity. Carrier removal at room temperature in p-type material is not influenced by the growth method or the dopant impurity. However, dopant effects are observed upon annealing.