液态铅铋氧浓度测量技术初步研究

液态铅铋合金是加速器驱动次临界系统(ADS)中散裂靶兼冷却剂的主要候选材料。氧浓度是影响液态铅铋合金(LBE)对结构材料腐蚀的关键因素,而氧传感器是实现液态铅铋合金中氧浓度精确测量的重要部件,本研究设计研制了一种液态铅铋系统氧传感器并基于自主研制的高温液态铅铋合金氧测控预研平台,初步开展了氧饱和LBE中的氧浓度测量实验。实验结果显示,300～400℃的氧饱和LBE中,氧传感器的电压信号(E)随温度(T)变化的实验曲线与理论曲线变化趋势相吻合;相对于300℃T350℃温度范围,氧传感器在350℃T400℃范围内的测量性能更好,仪器本身的系统误差约为17mV。     

液态铅铋合金流动速度场测量技术研究

液态铅铋合金是先进反应堆-加速器驱动的次临界系统(ADS)优选的靶材和冷却剂材料,液态铅铋流动速度场的测量是优化堆芯组件分布以及靶窗结构的一种重要手段。同时,冷却剂流动速度也是反映反应堆热工水力特征的重要参数之一。本文采用实验研究的方法,设计旋转搅动装置,通过对常温水与液态铅铋的流动速度场测量,验证了超声多普勒测速技术用于液态铅铋合金速度场测量的可行性。     

液态铅铋电磁流量计初步标定实验与分析

铅铋合金(Lead-Bismuth Eutectic,LBE)是加速器驱动次临界系统主选冷却剂材料之一,其热工流量测量面临高温、强腐蚀等苛刻环境。电磁流量计(Electromagnetic flow-meter,EMFM)是目前国际上用于铅铋流量测量的主选设备之一。研究发现,铅铋电磁流量计的标定技术对于其准确测量具有重要作用。本文基于PREKY铅铋技术预研实验平台,对自主研制的高温铅铋电磁流量计进行了标定实验与分析。在温度为350oC、流量为3.8–4.5m3?h-1的工况下,获得了液态铅铋电磁流量计的标定公式。标定公式计算值与实验值之间的误差范围是-4.9%–5.9%。     

液态铅铋合金热物性研究

液态铅铋合金(Lead-Bismuth Eutectic,LBE)是加速器驱动次临界系统(Accelerator driven subcritical,ADS)重要的散裂靶材料和冷却剂候选材料,其热力学物理性质是ADS研发过程中必须解决的基础问题。通过对已有定律的推算,得出液态铅铋合金熔沸点、密度、比热容、粘度、热导率的热物性公式;并拟合了其他学者的实验研究数据,得出计算铅铋物性的拟合公式。通过对比分析可知:拟合公式与已有定律推算公式趋向一致,吻合较好;且拟合公式更趋近实验值,精确度高,最大偏差不超过1%。     

液态铅铋合金综合实验装置初步设计

完成了铅铋合金综合试验装置（TCTL）的初步设计方案。 装置的总设计功率为500kW，为中等规模强迫流动的综合实验装置，由铅铋合金回路、除气系统、覆盖气体系统、冷却水系统、氧控制系统和数采和控制系统及供配电系统等组成。流动介质为纯度高于99．5％的铅铋合金，其中铅含量44．5％，铋含量55．5％。     

液态铅铋合金管内流动传热特性研究

为研究液态铅铋合金的流动传热特性,建立了相应的实验台架,并开展了传热特性试验,获得了1组针对液态铅铋合金的传热关系式。同时使用Fluent程序对圆管内铅铋合金的流动换热进行了模拟计算,以研究不同湍流普朗特数模型和不同湍流模型的选取对计算结果的影响。根据计算结果,探讨了不同湍流普朗特数模型对计算结果的影响,并得到了不同Peclet数下推荐使用的物理模型。     

包壳材料316Ti在液态铅铋中的腐蚀氧化层分析

本文利用自主研制的液态铅铋腐蚀试验装置,开展了中国铅基研究实验堆候选包壳材料316Ti在500℃氧饱和液态铅铋中3 000h的腐蚀试验。利用SEM及能谱仪对腐蚀后的样品进行了分析,结果显示316Ti样品腐蚀后形成了双层氧化膜,外层为疏松的Fe3O4氧化物,内层为致密的Fe-Cr尖晶石。随腐蚀时间的增加,外层氧化膜厚度变化不明显,但内层氧化膜逐渐增加,同时内层氧化膜沿晶界向基体生长,呈现出晶间腐蚀的特征。     

气泡在液态铅铋合金内上升行为的数值模拟

为利用气泡提升泵的概念设计来增加铅铋合金非等温回路的自然循环能力,需对充入回路的气泡大小进行选择。气泡的初始直径能显著影响气泡提升泵提升自然循环的能力。利用VOF模型,对不同初始直径的氦气泡在液态铅铋合金中的提升作用进行了数值模拟研究。结果表明,对于单个气泡,随着初始直径增大,提升能力呈先增加后趋于不变的趋势;初始直径较大的气泡易分裂生成较小的子气泡附着在管壁上,有影响传热效率的潜在可能。研究了在静止液态铅铋合金中,不同初始直径的氦气泡上升速度和上升高度与时间的关系及气泡分裂的原因。     

ADS中液态铅铋靶流动管道屏蔽计算

对于具有独立回路冷却液态铅铋(Liquid Lead-bismuth Eutectic,LBE)散裂靶的加速器驱动的次临界系统(Accelerator Driven Sub-critical System,ADS),高能质子束流辐照靶体产生的散裂产物进入管道后,会持续释放光子,会对靠近流动管道的工作人员造成辐射损伤。本文利用高能粒子输运程序Fluka计算散裂靶的散裂产物累积产额,然后利用放射性衰变计算程序Origen2计算散裂产物释放的光子强度及分群能谱,最后利用Fluka进行流动管道内光子源项的屏蔽计算。对一种典型的Y型LBE有窗靶进行的流动管道屏蔽计算结果表明,参照GB18871-2002职业性放射性工作人员年剂量当量限值20 m Sv的标准,若使用铅作为屏蔽材料,流动管道的管壁应该加厚20 cm。本文工作可为ADS系统LBE有窗靶的回路屏蔽设计提供参考。     

液态铅铋与水界面碎化行为的可视化实验

在中国铅基反应堆(CLEAR)换热器管道破口(SGTR)事故工况下,二回路高压水可能会直接与一回路铅铋共晶合金(LBE)接触,导致水/蒸汽混合物的急速沸腾,甚至发生蒸汽爆炸,从而危及反应堆的安全。为研究水与熔融LBE接触界面间的沸腾传热与蒸汽爆炸现象及机理,本文通过熔融LBE/水直接接触反应实验平台,依托高速摄像机记录熔融LBE入水爆炸或碎化过程。实验分析了LBE温度(250~500℃)、水温(25~80℃)对熔融LBE碎化行为的影响。结果显示,随着熔融LBE温度或水温的升高,LBE碎化质量中位粒径呈减小趋势;当熔融LBE与水接触界面温度大于水的均相成核温度时,蒸汽爆炸现象更易发生,碎化现象更明显。     

Metallurgical study on electro-magnetic flow meter and pump for liquid lead-bismuth flow

Corrosion and precipitation of corrosion products at an electrode of an electro-magnetic flow meter and in a narrow channel of an electro-magnetic pump (EMP) for lead bismuth (Pb-Bi) flow was investigated by means of metallurgical analysis. The roughened surface of inlet pump core at the entrance of EMP flow channel was observed, which indicated the occurrence of the so-called corrosion/erosion. In the same time, corrosion products were precipitated on the surface of the electrode and the outlet pump core. Localized corrosion/erosion and precipitation of corrosion products were observed by means of SEM/EDX analysis for their cross sections. The surface condition was different from each other depending on the place. The precipitation products were made of double-layers, that is, an outer layer and an inner layer. The inner layer (diameter range of 0–20 μm) contained Fe in the fraction of 99wt%, and the outer layer (diameter range of 20–40 μm) contained Pb-Bi in the fraction of 10wt%. The double layer structure might be caused by difference of precipitation period. These lumps of the precipitation products could be the cause of the channel plugging.     

组合阀流动阻力数值实验

组合阀的流动阻力是控制棒水压驱动系统设计的关键参数之一.利用计算流体力学软件 CFX-5.6 对组合阀进行数值模拟实验和三维流场计算,得到在不同差压下所对应的稳定流量,进而算出各流道阻力系数.通过比较流动阻力的计算结果,证实φ9阀的工作性能比φ6阀得到了提高.利用后处理工具分析φ9阀流道的流速、压力等流场分布特性,指出阻力损失的主要部分位于来水-侧壁-阀芯区这一段,并提出进一步的结构优化改进意见.     

Atomistic simulation of liquid lead and lead-bismuth eutectic

Lead and the Pb-Bi eutectic (Pb 55.9 at.%) have been modeled by a n-body potential derived from a second moment approximation of a tight binding Hamiltonian. The thermal behavior of the two systems in the liquid phase has been reproduced and relevant structural parameters have been evaluated and compared with experimental data. The diffusion coefficients and the activation energy for diffusion have been also evaluated.     

Critical temperature of the lead-bismuth eutectic (LBE) alloy

Liquid metals such as Bi and Pb and Pb-Bi eutectic alloy are serious contenders for use as coolant in LMFBRs in lieu of sodium due to a number of attractive characteristics (high density, low moderation, low neutron absorption and activation, high boiling point and poor interaction with water and air, etc.). Analysis of hypothetical accidents is of relevance to predict the catastrophe involving loss of coolant accident (LOCA) in LMFBRs. One key parameter to take into account is the critical temperature data of the liquid metals for reactor safety analysis. This communication reports the application of a theoretical model called internal pressure approach to predict the critical temperature (T_c) of the LBE alloy for the first time.     

High temperature oxidation of Fe-9Cr-1Mo steel in stagnant liquid lead-bismuth at several temperatures and for different lead contents in the liquid alloy

This research project deals with the feasibility studies concerning the construction of an hybrid reactor for the transmutation of long-lived radioactive wastes. In this context, the liquid lead-bismuth eutectic (LBE) is considered to be a good candidate for the spallation target material needed for the neutrons production necessary to the transmutation. In this hybrid reactor, the LBE, which is enclosed in a T91 (Fe-9%Cr) steel container, can induce corrosion concerns. If the oxygen content dissolved in Pb-Bi is higher than the needed content for magnetite formation, corrosion proceeds by oxidation of the steel. Previously, specific results were reported, obtained in stagnant liquid LBE at 470 °C. An analytical model taking into account the oxide layer structure has been carried out. It involves iron, oxygen and chromium bulk diffusion and diffusion via preferential paths such as liquid lead-bismuth nano-channels incorporated in the oxide layer structure and grain boundaries. In this paper, experimental results on corrosion kinetics, obtained at different temperatures with different percentages of lead in the lead-bismuth alloy, are presented. The model, adapted to the different experimental conditions, is compared to these kinetics and to experimental points coming from the literature at different temperatures in LBE, in pure lead and in pure bismuth.     

Fracture toughness assessment of ferritic-martensitic steel in liquid lead-bismuth eutectic

The presence of micro-cracks at the surface of a ferritic-martensitic steel is known to favour its embrittlement by liquid metals and thus decrease the mechanical properties of the structural materials. Unfortunately, conventional fracture mechanics methods cannot be applied to tests in liquid metal environment due to the opaque and conducting nature of the LBE. Therefore new methods based on the normalization technique for assessment of plain strain fracture toughness in LBE were examined. This paper discusses the assessment of the plain strain fracture toughness of T91 steel in liquid lead bismuth environment at 473 K, tested at a displacement rate of 0.25 mm min⁻¹ and makes the comparison with results obtained in air at the same temperature and displacement rate. Although there is a decrease of the fracture toughness by 20-30% when tested in LBE, the toughness of the T91 steel remains sufficient under the tested conditions.     

Dissolved oxygen control and monitoring implementation in the liquid lead-bismuth eutectic loop: HELIOS

A 12 m tall LBE coolant loop, named as HELIOS, has been developed by thermal-hydraulic scaling of the PEACER-300MWe. Thermo-hydraulic experiment and materials test are the principal purposes of HELIOS operation. In this study, an yttria stabilized zirconia (YSZ) based oxygen sensor that was hermetically sealed for long-term applications using the electromagnetically swaged metal-ceramic joining method, have been developed for high temperature oxygen control application over a long period of time. The rugged electrode design has been calibrated to absolute metal-oxide equilibrium by using a first principle of detecting pure metal-oxide transition using electrochemical impedance spectroscopy (EIS). During the materials tests in HELIOS, dissolved oxygen concentration was administered at the intended condition of 10⁻⁶ wt% by direct gas bubbling with Ar + 4%H₂, Ar + 5%O₂ and/or pure Ar while corrosion tests were conducted for up to 1000 h with inspection after each 333 h. During the total 1000 h corrosion test, oxygen concentration was measured by oxygen sensor. The result confirmed that the direct gas bubbling method is a viable and practical option for controlling oxygen concentration in large loops including HELIOS.     

Thermodynamic properties and equation of state of liquid lead and lead-bismuth eutectic

Since the 1950s, liquid lead (Pb) and lead-bismuth eutectic (Pb-Bi) have been studied in the USA, Canada and in the former-USSR as potential coolants for nuclear installations due to their very attractive thermophysical and neutronic properties. However, experimental data on the thermal properties of these coolants in the temperature range of interest are still incomplete and often contradictory. This makes it very difficult to perform design calculations and to analyse the normal and abnormal behaviour of nuclear installations where these coolants are expected to be used. Recently, a compilation of heavy liquid metal (HLM) properties along with recommendations for its use was prepared by the OECD/NEA Working Party on Fuel Cycle (WPFC) Expert Group on Lead-Bismuth Eutectic Technology. A brief review of this compilation and some new data are presented in this article. A set of correlations for the temperature dependence of the main thermodynamic properties of Pb and Pb-Bi(e) at normal pressure, and a set of simplified thermal and caloric equations of state for the liquid phase are proposed.     

铅铋堆嬗变燃料初步选型与分析

分离嬗变目前看来是次锕系核素回收和处理的比较可行路径,嬗变燃料研究也是第四代核能系统和先进核燃料循环技术的研究热点。本文简要归纳了嬗变燃料研究的主要特点以及当前国内外的研究现状,建议我国大力发展相关技术,特别是快堆嬗变燃料。此外,通过分析铅铋快堆/加速器驱动次临界系统(Accelerator Driven Sub-critical System,ADS)的特点和传统快堆燃料性质,得出氮化物嬗变燃料是目前铅铋堆嬗变燃料优选方案的结论。对中国科学院正在建设的ADS嬗变系统,期望未来考虑开展氮化物燃料的辐照和嬗变性能测试。     

High temperature wear and corrosion resistance of Co-free Ni-based alloy coatings on nuclear valve sealing surfaces

As a further step in obtaining high temperature wear and corrosion performances of a newly designed and laser cladded cobalt-free, nickel-based alloy coating (Ni-3) used for sealing surfaces of nuclear power valves, high temperature dry sliding wear behavior was characterized on a HT-1000 ball-on-disk tribometer at 360 °C, the hot corrosion test was conducted by plating Na₂SO₄ + K₂SO₄ mixed salt solution on the coating at 350 °C for 140 h. For comparison, the high temperature wear and corrosion resistance of the laser cladded Co-based Stellite06 and Fe-based Norem02 coatings were also investigated at the same testing conditions. It is found that the microhardness of the Ni-3 coating is about HV500 and close to Stellite06, but less than Norem02, the high temperature wear volume loss of Ni-3 coating is half of Norem02 and close to Stellite06. While during 350 °C hot corrosion test, the mass loss of Norem02 coating is the largest, Ni-3 coating is placed in the middle, Stellite06 coating shows the best hot corrosion resistance. Thus, the newly designed and laser cladded Ni-3 alloy coating appears to be the viable alternative material and technique for Co-free alloy, especially used in the nuclear valve sealing surfaces.