熔盐冷冻壁传热平衡数值模拟及实验研究

在熔盐堆燃料干法处理流程中,处理设备面临着严重的材质腐蚀问题。熔盐冷冻壁技术被视为保护相关设备耐受化学腐蚀的有效方法。熔盐冷冻壁厚度及温度场分布对防护效果及相关干法工艺参数有重要影响,在中心冷却式冷冻壁实验装置上采用FLi Na K熔盐介质开展了冷冻壁维持的实验研究,采用中心冷棒空气冷却的方式得到了不同壁温下冷冻壁的厚度范围及反应釜内温度场分布。结合数值模拟计算了冷冻壁的传热平衡工况,并与实验值进行对比和分析,得到了较为适宜的冷冻壁厚度的调节工况。冷冻壁厚度控制工艺和釜内温度场分布可为冷冻壁工艺容器的设计提供参考。     

熔盐冷冻壁形成及控制实验研究

乏燃料干法后处理中高温熔盐化学工艺过程中存在设备腐蚀问题,以凝固盐层作为容器保护内衬的熔盐冷冻壁技术被认为是一种可行的解决方式。为开展该技术研究,自行研制并搭建了一套硝酸盐(40.0-7.0-53.0wt%Na NO2-Na NO3-KNO3)冷冻壁技术研究实验装置。该装置上熔盐运行温度150-250oC,熔盐最大流量500 L·h-1,循环导热油运行温度5-120oC,导热油流量1.5-15 m3·h-1。目前在该装置上开展了冷冻壁静态形成及平衡维持等工艺研究,实验中采用容器外壁循环导热油冷却换热实现冷冻壁的形成及维持,并试验了冷冻壁技术在熔盐静态工况下应用的工艺条件。冷冻壁形成平均速率可控制在0.2-0.5 mm·min-1。在冷冻壁静态形成过程中,随厚度增大,热交换量逐渐减小,同时冷冻壁层温差逐渐增大,并均呈衰减趋势变化;处于平衡维持状态时,径向温度分布、热流量及冷冻壁厚度均保持稳定,熔盐发热功率即为平衡状态时的热流量,其大小同时与外壁导热油的冷却热流量相等;实验还获得了较理想的静态应用工艺操作条件,为氟化物熔盐冷冻壁的研究及应用积累了经验。     

熔盐电解法乏燃料干法后处理技术研究进展

熔盐电解法是目前最有前途的干法后处理技术,适合于处理氧化物和金属等不同类型乏燃料。熔盐电解法主要包括四个核心流程,即首端处理、电解还原、电解精炼和提取以及废物处理。本文以国际上最新的研究进展为蓝本,综述熔盐电解法乏燃料后处理技术的基本流程以及待解决的关键问题。     

乏燃料干法后处理中铀电沉积行为模拟研究

基于多物理场耦合软件COMSOL,在LiCl-KCl熔盐体系中建立以电解槽为阳极、双石墨棒为阴极的氯化铀电沉积行为三维数值模型。通过研究熔盐中铀离子的运动情况以及对阴极几何形状的实时计算,得到了铀电沉积的沉积层厚度随时间的变化情况,得到了铀离子阴极沉积行为与阴极表面的位置、熔盐铀离子浓度、反应温度、平均电流密度之间的关系。研究中,还将模拟结果与氯化铀熔盐电解实验数据进行对比,计算与实验结果拟合良好,证明铀电沉积行为模拟的可靠性,模拟结果可为乏燃料干法后处理中铀的提取提供设计参考。      

干法后处理技术典型流程综述

虽然基于溶剂萃取的Purex流程在乏燃料后处理几十年的应用中取得的成功,使得水法后处理至今没有发展出可以取代这一流程的新萃取剂,但干法后处理却有了两种可供进一步发展的流程:氟化物挥发法和高温电化学法。氟化物挥发法存在的最大问题是热力学上PuF6必须在有大量F2过剩的条件下才稳定。高温电化学法适合于处理合金元件,以及氧化物和碳化物元件。首先,将核燃料熔解在熔盐中,然后,电解使铀钚在阴极上沉积,再对阴极上沉积出来的铀钚进行精制而得到铀钚产品。但该方法存在熔盐对MOX的熔解能力和对过程设备的腐蚀问题。     

俄罗斯氧化物乏燃料电沉积流程研究进展

高温熔盐干法后处理以熔盐作为电解质,通过电解精炼和电沉积回收核燃料中的铀和钚。目前,俄罗斯、美国、日本、韩国和欧盟等国均在积极发展乏燃料高温熔盐干法后处理技术的研究,其中俄罗斯的金属氧化物核燃料电沉积流程是经典的流程之一。本文对俄罗斯原子反应堆研究所(Research Institute of Atomic Reactors, RIAR)发展的氧化物乏燃料高温熔盐电沉积干法后处理的发展现状、流程及特点进行了综述。     

俄罗斯氧化物乏燃料电沉积流程研究进展

高温熔盐干法后处理以熔盐作为电解质，通过电解精炼和电沉积回收核燃料中的铀和钚。目前，俄罗斯、美国、日本、韩国和欧盟等国均在积极发展乏燃料高温熔盐干法后处理技术的研究，其中俄罗斯的金属氧化物核燃料电沉积流程是经典的流程之一。本文对俄罗斯原子反应堆研究所（Research Institute of Atomic Reactors,RIAR）发展的氧化物乏燃料高温熔盐电沉积干法后处理的发展现状、流程及特点进行了综述。     

干法后处理流程中电解精炼设备研发进展北大核心CSCD

高燃耗快堆乏燃料具有高钚含量、强放射性、高释热率等特点。基于溶剂萃取原理的水法后处理工艺存在溶剂易辐解等问题,宜对高燃耗快堆乏燃料采用干法后处理工艺进行处理。熔盐电解干法工艺采用耐辐照的无机盐为介质,通过电化学方法分离回收锕系元素,是最具应用前景的干法后处理技术。在熔盐电解干法工艺流程中,承担锕系元素分离任务的电解精炼单元是核心环节。本文调研了乏燃料干法后处理过程中电解精炼设备的研发进展,分析了电解精炼设备关键技术和发展趋势,为我国快堆乏燃料电解精炼设备的研发提供了参考。     

镧系及锕系元素在离子液体中的电化学行为

乏燃料回收是核燃料循环的核心,对核安全和核能可持续发展具有重要的意义,其分为使用水溶液的湿法和不使用水溶液的干法处理。熔盐电解技术是乏燃料干法回收的重要方法之一,但其工艺温度往往在数百摄氏度,对设备和能耗要求都很高。离子液体具有电化学窗口宽、低熔点、低蒸汽压、热稳定性好等优点,有望替代高温熔盐用于乏燃料干法回收。本文概述了镧系元素和锕系元素在离子液体中电化学方面的研究状况,表明离子液体用于乏燃料干法回收是可行的,但需要更多的基础性研究。     

高温熔盐中氧化物乏燃料的电化学还原研究进展

干法后处理流程可应用于快堆乏燃料后处理。由美国开发的熔盐电解精炼流程是目前最具应用前景的干法后处理流程之一。为了将电解精炼流程应用于氧化物乏燃料后处理,需要将氧化物乏燃料转化为金属。目前电化学还原是应用最广的氧化物乏燃料还原方法,但是该过程仍然存在亟待解决的关键科学与技术问题。本文针对氧化物乏燃料电化学还原研究进展进行综合阐述,主要包括过程简介、研究现状及电化学还原机理等几个方面。     

The protective performance of a molten salt frozen wall in the process of fluoride volatility of uranium

The fluoride volatility method(FVM) is a technique tailored to separate uranium from fuel salt of molten salt reactors. A key challenge in RD of the FVM is corrosion due to the presence of molten salt and corrosive gases at high temperature. In this work, a frozen-wall technique was proposed to produce a physical barrier between construction materials and corrosive reactants.The protective performance of the frozen wall against molten salt was assessed using FLiNaK molten salt with introduced fluorine gas, which was regarded as a simulation of the FVM process. SS304, SS316 L, Inconel 600 and graphite were chosen as the test samples. The extent of corrosion was characterized by an analysis of weight loss and scanning electron microscope studies. All four test samples suffered severe corrosion in the molten salt phase with the corrosion resistance as: Inconel 600 SS316 L graphite SS304. The presence of the frozen wall could protect materials against corrosion by molten salt and corrosive gases, and compared with materials exposed to molten salt, the corrosion rates of materials protected by the frozen wall were decreased by at least one order of magnitude.     

Corrosion study of a highly durable electrolyzer based on cold crucible technique for pyrochemical reprocessing of spent nuclear oxide fuel

The application of the cold crucible technique to a pyrochemical electrolyzer used in the oxide-electrowinning method, which is a method for the pyrochemical reprocessing of spent nuclear oxide fuel, is proposed as a means for improving corrosion resistance. The electrolyzer suffers from a severe corrosion environment consisting of molten salt and corrosive gas. In this study, corrosion tests for several metals in molten 2CsCl–NaCl at 923 K with purging chlorine gas were conducted under controlled material temperature conditions. The results revealed that the corrosion rates of several materials were significantly decreased by the material cooling effect. In particular, Hastelloy C-22 showed excellent corrosion resistance with a corrosion rate of just under 0.01 mm/y in both molten salt and vapor phases by controlling the material surface at 473 K. Finally, an engineering-scale crucible composed of Hastelloy C-22 was manufactured to demonstrate the basic function of the cold crucible. The cold crucible induction melting system with the new concept Hastelloy crucible showed good compatibility with respect to its heating and cooling performances.     

Natural convection heat transfer test for in-vessel retention at prototypic Rayleigh numbers – Results of COPRA experiments

Large-scale COPRA experiments were performed to investigate the natural convection heat transfer in melt pools for the in-vessel retention during severe accidents in Chinese large-scale advanced PWRs. Both water and binary mixture of 20 mol% NaNO₃ – 80 mol% KNO₃ were used as the melt simulant material in performed tests. Due to the full scale geometry of the COPRA test section, the Rayleigh numbers of the melt pool could reach up to the prototypic magnitude of 10¹⁶. Natural convection heat transfer tests at prototypic Rayleigh numbers have been performed to study the influence of the heat generation rate and melt simulant material on the melt pool temperature, heat flux distribution and heat transfer capability. The comparisons of the melt pool temperature and heat flux distribution from water experiments and molten binary salt experiments showed that the crust formation along the inner surface of the vessel wall could impact the heat transfer characteristics of the melt pool. And the heat flux distribution from COPRA water tests and molten salt tests were in good agreement with those from Jahn-Reineke water experiments and RASPLAV molten salt experiments, respectively. The heat transfer capability of the melt pool Nu_dn from COPRA molten salt tests were larger than those from water tests, but both were lower than those from ACOPO and BALI predictions within the same range of Rayleigh numbers (10¹⁵ – 10¹⁷).     

新概念熔盐堆非能动余热排出系统中钠热管的特性研究

高温钠热管作为一种高效的换热装置，在导热性、等温性以及非能动特性方面具有显著的优点。将高温钠热管应用到新概念熔盐堆非能动余热排出系统中具有重要的研究意义。本工作通过数值方法研究了高温钠热管在熔盐堆事故工况下的瞬态运行特性。钠热管的物理模型主要包括管壁、吸液芯及蒸汽区3个耦合传热区域。通过对上述3个区域建立合理的数学模型并采用有限元的方法，利用 FORTRAN进行编程，最终得到高温钠热管启动过程中的温度、速度、压力分布。结果表明：熔盐堆事故状态下，钠热管从启动到稳态过程中其运行特性良好且具有很高的传热效率。     

影响CaCl2-NaCl熔盐电脱氧法制备金属Zr的工艺条件

将氧化物转化为金属是熔盐电解精炼干法后处理氧化物乏燃料流程的关键步骤之一。在等摩尔CaCl₂-NaCl混合熔盐体系中,以石墨棒为阳极,采用高温烧结后的ZrO₂模拟UO₂开展了电脱氧制备金属Zr的FFC剑桥工艺条件优化。研究了工艺条件（槽电压、电解时间、烧结温度和电解温度等）对电脱氧制备Zr的影响。采用场发射扫描电子显微镜(SEM)和X射线衍射(XRD)分别分析了电解前后ZrO₂阴极的微观结构和物相组成。优化后的工艺条件为：电压3.4 V、电解时间12 h、烧结温度900 ℃和电解温度722 ℃。同时,研究结果表明, ZrO₂电脱氧还原为Zr时,存在中间产物CaZrO₃和ZrO。     

硬壳对熔融池换热特性影响试验研究

基于大型熔融池换热特性试验台架COPRA,分别采用水和非共晶摩尔比例20%NaNO₃-80%KNO₃混合物作为熔融物模拟物,进行熔融池换热特性试验研究。熔融池瑞利数Ra′可达到10¹⁶量级,与反应堆真实情况下的量级一致。试验对比了水试验和熔盐试验得到的熔融池温度场分布和壁面热流密度分布。结果表明,熔融物硬壳的形成对熔融池换热特性产生了明显影响。试验拟合得到了熔融池换热特性关系式,其中热流密度关系式与国际上其他试验得到的结果符合得较好。在同等Ra′量级下,COPRA试验得到的熔融池向下封头壁面的传热量较国际上其他试验的结果低。     

Corrosion behavior of Hastelloy-N alloys in molten salt fluoride in Ar gas or in air

The effects of air on the corrosion of Hastelloy-N alloys in molten salt coolant containing fission product elements were investigated to determine the safety of structural materials in high-temperature reactors cooled with fluoride salt. Corrosion tests of Hastelloy-N in the molten fluoride salt FLiNaK in an alumina crucible and a graphite crucible under argon gas or air were performed at 773–923 K for 100 h. The depth of corrosive attack, as well as the extent of chromium and molybdenum depletion, increased with increasing temperature. The extent of Hastelloy-N corrosion in molten salt under air was significantly greater than under argon gas. The effect of adding the impurity cesium iodide to molten salt containing nuclear waste fuel on the corrosion behavior was negligible.     

Corrosion Phenomenon of Stainless Steel in Boiling Nitric Acid Solution Using Large-Scale Mock-Up of Reduced Pressurized Evaporator

To evaluate the component life in a spent nuclear fuel reprocessing plant, a large-scale mock-up test apparatus of a reduced pressurized thermosiphon evaporator was constructed, and the corrosion mechanism of a heat transfer tube made of ultralow carbon type 304 stainless steel in boiling nitric acid solution was studied. The corrosion tests were conducted for about 36,000 h, and changes in the corrosion amount and rate in the test duration were discussed. The relationships between the amount of corrosion and tube surface temperature and heat flux were investigated, and the corrosion propagation mechanism considering intergranular penetration was studied based on the observations of morphologies of corrosion surfaces and the measurements of intergranular penetration depths. After a long duration, the increases in the corrosion amount and rate saturated when intergranular penetration and grain dropping occurred by turns. This result means that a linear estimation can be applied to the life prediction for corrosion. Three portions of the tube were observed, and the amounts of corrosion were different among the three portions, but no difference in the morphology of intergranular corrosion existed. The amount of corrosion was affected by both tube surface temperature and heat flux. A large amount of corrosion could be observed in both the boiling starting portion and the top, where high tube surface temperature and heat flux were observed.