差动变压器式单点液态金属液位计的研制

研制了一种全新的差动变压器(LVDT)式单点液态金属液位计,并对其性能进行了测试。结果表明,该型液位计不受诸多误差因素的影响(如温度效应),其探头可在0~650℃范围内长期工作,其对液面测量误差优于±2mm。该型液位计配合专用的驱动机构,可实现液面的连续测量。     

永磁式钠流量计的研制

研制了量程为5m~3/h和0.5m~3/h的两种永磁式钠流量计,介绍了它们的结构、分度特性与阻力特性的理论计算、标定试验及误差分析,给出了分度特性的解析表达式。     

高精度钠压力传感器的研究

本文研究了一种以测力环为敏感元件的高精度钠压力传感器,这种传感器具有良好的线性度和微小的变差,经温度效应在线修正后,可获得高精度的钠压力测量。     

高精度钠压力传感器的研究

本文研究了一种以测力环为敏感元件的高精度钠压力传感器。给出了它的结构图,介绍了测力环的计算,应变计的安装,测量线路和分度检定装置,提出了温度效应在线修正的方法,进行了检定误差分析。通过比较,表明这种传感器具有良好的线性度和微小的变差。通过温度效应的在线修正,可以得到高精度的钠压力测量数据。     

高精度钠压力传感器的研究

本文研究了一种以测力环为敏感元件的高精度钠压力传感器。给出了它的结构图,介绍了测力环的计算,应变计的安装,测量线路和分度检定装置,提出了温度效应在线修正的方法,进行了检定误差分析。通过比较,表明这种传感器具有良好的线性度和微小的变差。通过温度效应的在线修正,可以得到高精度的钠压力测量数据。     

钠冷阱丝网布置的最优化分析

本文主要讨论冷阱中丝网布置的优化问题。对于冷阱中的质量传递问题采用了一维模型，对丝网沉积面积随时间而变的因素进行了简化处理，提出了液态钠流经丝网区一次后应尽量满足对杂质浓度的要求，且在整个丝区杂质析出均匀，丝网用量最小情况下的最优化模型。利用诚特里亚金极大值原理对所得模型在等温冷阱情况下进行了求解，并对非等温冷阱的情况进行了性分析。所得结论可为工程设计提供理依据。     

雾状钠火钠滴燃烧比率的模拟及应用

将雾状钠火中钠滴的燃烧分成预燃阶段和燃烧阶段,利用雾状钠火程序计算得到钠滴燃烧比率和时间的关系曲线,分别用幂函数、指数函数和线性函数对曲线进行拟合,拟合效果较好。拟合函数中包含钠滴下落时间和钠滴最大燃烧比率等参数,这些参数可通过钠滴下落燃烧试验或雾状钠火程序计算得到。通过推导得到了雾状钠火燃烧和单个钠滴燃烧的关系,钠滴燃烧比率的拟合函数被用来模拟雾状钠火燃烧的过程,包括用于计算已燃烧的钠质量、空气中未燃烧的钠质量、进入钠池的钠质量和雾状钠火的燃烧速率。当雾状钠火燃烧过程中钠泄漏流量恒定不变时,空气中未燃烧的钠质量和钠泄漏流量呈正比,雾状钠火的燃烧速率和钠泄漏流量呈正比。雾状钠火的燃烧速率和钠火造成的事故工艺间内的温度与压力变化直接相关。雾状钠火的燃烧速率被用来求解钠气溶胶的生成速率、钠燃烧火焰层和空气之间的传热、钠燃烧火焰层和墙壁之间的传热。总之,使用简单的函数模拟钠滴的燃烧比率曲线,将雾状钠火燃烧当成事故工艺间的热源和钠气溶胶源作为输入,便可模拟雾状钠火的整个燃烧过程,计算得到工艺间温度、压力和钠气溶胶浓度的变化。钠滴的燃烧比率曲线、雾状钠火的燃烧速率曲线还可与试验数据进行对比验证后作为雾状钠火模拟的输入,这种模拟方法可用于钠火事故安全分析中雾状钠火的模拟。     

硅酸钙与钠及钠蒸气的相容性研究

中国实验快堆主容器上方旋塞采用硅酸钙制品作为绝热材料,该材料存在与钠蒸气以及液态钠接触的可能性。通过开展硅酸钙绝热材料在高温钠、钠蒸气环境中的腐蚀实验,并对实验后样品进行表面形貌、剖面形貌以及导热性能测量等观察分析,以评估该材料与高温钠和钠蒸气环境的相容性能。结果表明:硅酸钙在液态钠和钠蒸气中不发生激烈的化学反应;液态钠和钠蒸气对硅酸钙的作用仅发生在表层,硅酸钙材料内部的颜色及形貌与试验前没有区别,未见钠的渗入,也未见任何裂纹或破损;试验后硅酸钙材料的热导性能与试验前基本相同,液钠和钠蒸气对硅酸钙的绝热性能基本没有影响。硅酸钙表现出较好的抗钠腐蚀性能。     

YSJ-1型油水界面检测仪的研制与应用

YSJ-1型油水界面检测仪是一种用核物理方法检测油水分离罐中油水分界面高度和液面总高度的新型仪表。它采用了非线性拟合、放射源衰变自动补偿、介质温度一密度自动补偿等多项技术，具有非接触测量、抗结蜡和维修量小等常规仪表不可比拟的优点。测量范围：液面总高度0-200cm，油水分界面高度0—l00cm；测量精度：液面总高度为l％，油水分界面高度为2％；反应时间：≤l0s；长期稳定性：≤0．5％量程／48h；温漂：≤0．01％量程／℃。该检测仪主要应用于油田生产中油水分离罐的油水分界面高度和液面总高度的检测，也可用于其它场合存储罐中两种不同液体的分界面高度和液面总高度的检测。     

多路二线制料位计

介绍了多通道二线制料位计的设计,与传统料位计相比,信号在长距离传输中不衰减、波形不畸变、抗干扰能力强,节省传输电缆芯线数目,也节省主机资源。     

减小核料位计统计涨落的方法

本文介绍了一种减小核料位计统计涨落的方法,该方法应用到核料位计后,在同等工业现场工作环境下可减小核料位计使用放射源的活度和防止控制过程的抖动.对于节约成本,减少放射性环境污染具有重要的意义.该方法还可应用到其它核仪表中,具有广泛的应用价值.     

薄膜X射线测厚仪的研制

介绍了用于薄膜测厚的全自动往复式X射线测厚仪的研制,详述了该测厚仪的硬件组成和测厚的工作原理.该装置已成功运行在双向拉伸聚丙烯(BOPP)薄膜生产线上.     

多通道γ射线物位计与可编程序控制机(PC)联机装置在立式水隔离灰浆泵上的应用

对立式水隔离灰浆泵的信号监测、控制系统所采用的多通道γ射线物位计与ＰＣ机联网装置的工作原理、应用实例、电路组成、监控程序及使用效果等作了介绍与讨论。     

Space reactor power systems with no single point failures

Nuclear reactor power systems could revolutionize space exploration and support human outpost on the moon and Mars. This paper reviews various energy conversion technologies for use in space reactor power systems and provides estimates of the system's net efficiency and specific power, and the specific area of the radiator. The suitable combinations of the energy conversion technologies and the nuclear reactors, classified based on the coolant type and cooling method, for best system performance and highest specific power, are also discussed. In addition, a number of power system concepts with both static and dynamic energy conversion, but with no single point failures in reactor cooling, energy conversion and heat rejection, and for nominal electrical powers up to 110 kW_e, are presented. The first two power systems employ reactors cooled with lithium and sodium heat pipes, SiGe thermoelectric (TE) and alkali-metal thermal-to-electric conversion (AMTEC), and potassium heat pipes radiators. The reactors heat pipes operate at a fraction of the prevailing capillary or sonic limit, and in the case of a multiple heat pipes failure, those in the adjacent modules remove the additional heat load, thus maintaining the reactor adequately cooled and the power system operating at a reduced power. The third power system employs SiGe TE converters and a liquid metal cooled reactor with a divided core into six sectors that are neurotically and thermally coupled, but hydraulically decoupled. Each sector has a separate energy conversion loop, a heat rejection loop, and a rubidium heat pipes radiator panel. When a core sector experiences a loss-of-coolant, the fission power of the reactor is reduced, and that generated in the sector in question is removed by the circulating coolant in the adjacent sectors. The fourth power system employs a gas cooled reactor with a core divided into three identical sectors, and each sector is coupled to a separate Closed Brayton Cycle (CBC) loop with He-Xe binary mixture (40 g/mol) working fluid, a secondary loop with circulating liquid Nak-78, and two water heat pipes radiator panels.     

ZHM—1型智能核密度计的研制与应用

介绍一种新型智能核密度计，它采用了对被测介质密度非线性拟合，放射源衰变自动补偿，介质温度－密度自动补偿等多项新技术，与国内同类仪表相比，具有测量精度高，操作简便，自检等功能。     

Regional and world level scenarios for sodium fast reactor deployment

The revival of interest towards fast neutron reactors is due to their capability of producing more fissile material than they consume and of reducing the burden on the disposal of high-level waste containing very-long life nuclides.Several studies on the impact of transition scenarios have been launched by international and national organizations. They show that the relevance of transitioning to fast reactors depends on the expected future role of nuclear energy and on the development of advanced nuclear technologies.The purpose of the present study is to investigate the maximum deployable capacity of sodium-cooled fast reactors at regional and world levels mainly from the viewpoint of natural uranium and fissile plutonium availability for presumed energy demand projections. The calculations have been performed with the DESAE code.The results show that the maximum deployable capacity of sodium-cooled reactors at regional level is not limited by the plutonium and uranium availability with a reprocessing capacity of 3000 t/yr, whereas for a global scenario it strongly depends on the availability of natural uranium resources and plutonium stockpiles.     

Development of analysis method for sodium void reactivity of step type FBR cores by using group-wise Monte Carlo code

A new method is proposed to separate the sodium void reactivity of step type FBR cores to components including non-leakage terms and a leakage term by using a newly developed perturbation code MCPERT where fluxes and adjoint fluxes are derived from a group-wise Monte Carlo code. The step type FBR core is a core where the height of the inner core is smaller than that of the outer core and a large sodium plenum region is located above the core so as to decrease the sodium void reactivity. The conventional diffusion perturbation method cannot treat such a large void region due to the diffusion approximation, while the Monte Carlo code can treat it exactly. In this study, a group-wise Monte Carlo code GMVP with a 70-group constant set JFS-3-J3.3 is employed to evaluate the neutron fluxes and adjoint fluxes which are used as inputs to the MCPERT code to evaluate the non-leakage terms. The leakage term is derived from the difference of the total sodium void reactivity evaluated by the direct calculation of GMVP and the summation of the non-leakage terms. It is found that the proposed method can provide the result approximately consistent to the ratio of the reactivity components derived from the conventional method.