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.     

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.     

高温原位氟化熔盐红外吸收光谱装置研制

钍基熔盐堆是第四代核能系统的候选堆型之一,熔盐由于优异的传热性能与中子特性,被作为冷却剂和燃料盐载体在堆内运行,其微观结构对其物理化学性质有重大影响,因此,研究熔盐结构对熔盐制备净化、腐蚀控制以及熔盐堆的设计、建造和安全运行都有重要的指导作用。红外吸收光谱(Infrared Absorption Spectroscopy,IR)是研究熔盐结构的有力工具之一,但标准仪器无法实现高温(500oC)熔盐的测量。本研究在解决了样品加热、环境气氛控制以及腐蚀等难题后,研制了一套适用于高温氟化熔盐实验的高温原位红外吸收光谱装置。该装置中,加热炉为左右两开式,可以直接放进标准红外光谱仪样品仓内;样品池为组装式,主体是能够耐氟化熔盐腐蚀的镍基哈氏合金,窗片为单晶Si C或金刚石;样品池整体呈倒"T"型,上端密封盖带有进气口和出气口,可实现抽真空或通惰性气体的操作。该装置可以实现25-600oC温度范围测量,波段范围覆盖近红外和中红外(14 700-400 cm~(-1))。通过使用常温水和高温亚硝酸钠的红外光谱实验对该装置的可靠性进行了验证。利用该装置,我们成功地获得了高温氟锂铍(FLi Be)熔盐的红外吸收光谱。     

金属铈在NaCl-KCl和NaCl-KCl-CeCl3熔盐体系中的溶解行为

通过离子浓度测定、气体生成法、熔盐耗酸量测定、X射线衍射和差重分析法,综合评判了金属铈在NaCl-KCl和NaCl-KCl-CeCl₃熔盐体系中的溶解行为。结果表明,金属铈在熔盐中的损失存在物理溶解和化学反应,可生成氧化物、氮氧化物、氯氧化物等,同时还可与坩埚成分发生反应。金属铈及其化合物在熔盐中的溶解度很低,但会生成新相,新相中含有大量的铈。采用嵌套式坩埚并增加内部小坩埚高度,可有效降低金属铈的溶解损失量。     

Dissolution Behavior of Cerium in NaCl-KCl and NaCl-KCl-CeCl3 Molten Salt System

The dissolution behavior of cerium in NaCl-KCl and NaCl-KCl-CeCl₃ molten salt systems was evaluated by ion concentration determination, gas generation method, determination of acid consumption of molten salt, X-ray diffraction and differential gravimetry analysis method. The results show that the loss of cerium in molten salt involves physical dissolution and chemical reactions, which can produce oxides, nitrogen oxides, chlorine oxides and other forms. The cerium can also react with the composition of crucible. Cerium and its compounds have a low solubility in molten salts, but new phases are formed, which contain large amounts of cerium. The dissolution loss of cerium can be effectively reduced by using nested crucible and increasing the height of small crucible.     

High-temperature corrosion characteristics of yttria-stabilized zirconia material in molten salts of LiCl-Li2O and LiCl-Li2O-Li

The isothermal and cyclic corrosion behavior of yttria (Y₂O₃)-stabilized zirconia (ZrO₂) in a LiCl-Li₂O molten salt were investigated at 650 °C in an argon atmosphere. During isothermal and cyclic corrosion tests in the molten salt of LiCl-Li₂O for 168 h and 7 thermal cycles, the corrosion rate was very low, whereas under the molten salt of LiCl-Li₂O-Li for 168 h, the corrosion rate was almost 10 times higher than that in the molten salt of LiCl-Li₂O. No corrosion product was detected until 168 h for the isothermal corrosion test, however, after 7 thermal cycles, a very-low-intensity Li₂ZrO₃ peak was detected at the beginning stage of the chemical reaction between ZrO₂ and Li₂O. Additionally, in the molten salt of LiCl-Li₂O-Li for 168 h, a large amount of Li₂ZrO₃ was formed, with evidence of marked cracks, pores, and spallations on the corroded surface. The introduction of Y₂O₃-stabilized ZrO₂ was beneficial in increasing the hot corrosion resistance of the structural materials used to handle molten salts containing Li₂O at elevated temperature without forming a lithium at the cathode during the electrolytic reduction process.     

一体化小型氟盐冷却高温堆瞬态特性分析

根据下一代核能系统的发展目标,提出了采用自然循环的一体化小型氟盐冷却高温堆的概念。利用修改后的RELAR5-MS系统分析程序,建立了一体化小型氟盐冷却高温堆模型,并得到其稳态特性参数。在此基础上,对其在满功率运行状态下的反应性引入事故和失热阱事故进行了分析。分析计算表明,在反应性事故工况下,由于自然循环的存在,堆芯冷却剂流量随着堆芯温度发生动态变化,最终达到新的稳态,燃料棒和冷却剂温度均处于安全限值范围内。在失热阱事故下,反应堆负反馈的特性使得堆芯功率逐渐降低并实现自动停堆,即使不考虑余热排出系统的作用,燃料组件和冷却剂温度上升缓慢,在140 h内,燃料棒和冷却剂温度均处于全限值范围内。结果表明,一回路采用自然循环冷却的一体化小型氟盐冷却高温堆具有良好的固有安全性。     

1 GWt PB-FHR负荷跟踪运行特性分析

为研究熔盐堆系统在商业应用中的价值,分析其是否满足电网负荷的变化需求和安全运行的能力,本文以1 GWt球床式氟盐冷却高温堆（PB-FHR）为研究对象,仿真计算其在负荷跟踪模式下的瞬态行为和运行特性。以RELAP5/MOD4.0程序为研究工具,并植入相关的熔盐物性与计算关系式,建立氟盐冷却高温堆的热工水力系统与功率控制系统的仿真模型,对典型负荷工况参数变化情况下控制系统的响应特性进行仿真分析。结果表明：该氟盐冷却高温堆系统在设计的控制逻辑的调控下,展示出良好的负荷跟踪运行能力,堆芯功率能迅速响应负荷变化,功率超调和温度超调小,反应堆的运行参数始终处于合理的运行范围内。     

ADS熔盐散裂靶中子学性能研究

与传统加速器驱动次临界系统（ADS）采用金属靶作为散裂中子靶的设计不同,加速器驱动次临界熔盐堆（AD-MSRs）采用靶堆一体的设计,直接使用燃料熔盐作为散裂中子靶。由于熔盐靶的中子学性能直接影响AD MSRs的能量放大系数、核废物的嬗变和核燃料增殖的效率,所以本研究基于MCNPX程序,详细计算了高能质子轰击氟盐和氯盐两种熔盐靶产生的散裂中子产额、散裂中子能谱、能量沉积分布以及散裂产物等中子学性能,并与液态Pb和铅铋共熔体（LBE）两种液态金属靶进行了对比。计算结果表明,熔盐靶在散裂中子产额上与液态金属靶有一定的差距,但熔盐靶内能量沉积分布的梯度较小,更有利于靶区的热量导出。与液态Pb和LBE靶相比,熔盐靶的散裂产物中包含更多的气体以及高质量数的α发射体核素。     

Corrosion characteristics of Hastelloy N alloy after He+ ion irradiation

With the goal of understanding the invalidation problem of irradiated Hastelloy N alloy under the condition of intense irradiation and severe corrosion, the corrosion behavior of the alloy after He⁺ ion irradiation was investigated in molten fluoride salt at 700 °C for 500 h. The virgin samples were irradiated by 4.5 MeV He⁺ ions at room temperature. First, the virgin and irradiated samples were studied using positron annihilation lifetime spectroscopy (PALS) to analyze the influence of irradiation dose on the vacancies. The PALS results showed that He⁺ ion irradiation changed the size and concentration of the vacancies which seriously affected the corrosion resistance of the alloy. Second, the corroded samples were analyzed using synchrotron radiation micro-focused X-ray fluorescence, which indicated that the corrosion was mainly due to the dealloying of alloying element Cr in the matrix. Results from weight-loss measurement showed that the corrosion generally correlated with the irradiation dose of the alloy.     

温度与气压对LiF蒸发速率的影响

减压蒸馏技术是钍基熔盐反应堆燃料处理中回收和纯化核燃料载体氟盐(7LiF-BeF2)可选的技术。本工作使用自行设计并研制的热失重蒸发炉,通过观察燃料载体盐的主要成分LiF在减压蒸馏过程中的失重行为,研究了蒸发温度和气压对LiF蒸发速率的影响。结果表明:LiF的蒸发速率在随着蒸发温度的升高以及气压下降而增大的变化中存在着明显的拐点,蒸发速率的突变可以用熔盐在低气压条件下的沸腾来解释。并进一步讨论了熔盐的沸腾对减压蒸馏回收和纯化燃料载体氟盐的影响。     

Chemistry and technology of Molten Salt Reactors - history and perspectives

Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous ‘on-line’ reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R&D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium.     

The feasibility of using molten carbonate corrosion for separating a nuclear surrogate for plutonium oxide from silicon carbide inert matrix

Silicon carbide is one of the prime candidates as a matrix material in inert matrix fuels (IMF) being designed to reduce the plutonium inventories. Since complete fission and transmutation is not practical in a single in-core run, it is necessary to separate the non-transmuted actinide materials from the silicon carbide matrix for recycling. In this work, SiC was corroded in sodium carbonate (Na₂CO₃) and potassium carbonate (K₂CO₃), to form water soluble sodium or potassium silicate. Separation of the transuranics was achieved by dissolving the SiC corrosion product in boiling water. Ceria (CeO₂), which was used as a surrogate for plutonium oxide (PuO₂), was not corroded in these molten salt environments. The molten salt depth, which is a distance between the salt/air interface to the upper surface of SiC pellets, significantly affected the rate of corrosion. The corrosion was faster in K₂CO₃ than in Na₂CO₃ molten salt at 1050 °C, when the initial molten salt depths were kept the same for both salts.     

FLiNaK熔盐中CsF的蒸发与分离

Cs的同位素是核裂变的主要产物之一,在熔盐反应堆液态燃料盐中以Cs F的化学形态存在,定期从燃料盐中除去或减少其含量将有助于提高反应堆的中子经济性。本文用FLi Na K熔盐模拟熔盐堆载体盐FLi Be体系,研究了Cs F在不同蒸发条件下的蒸发行为,并尝试进行了减压蒸馏和金属Li还原蒸发技术分离Cs F的实验研究。研究表明,在5 Pa蒸馏压力下,Cs F的蒸发量随温度呈线性上升趋势,780oC时Cs F的含量由1%降到0.14%,分离率达86%,但此时载体盐的蒸发量达9.5%;在常压、700oC条件下,熔盐中Cs F的蒸发比例随还原剂Li的添加量而提高,当添加的金属Li的摩尔浓度与Cs F为120:1时,Cs F分离率达91%。研究结果为了解Cs F在氟盐体系中的蒸发行为和建立可行的分离方法提供基础实验依据。     

随机球床模拟方法对氟盐冷却高温堆中子学参数的影响分析

氟盐冷却高温堆（FHR）采用氟盐冷却球形燃料元件,其中子物理计算面临双重不均匀性问题：燃料球在堆芯内的随机排布和包覆燃料颗粒在燃料球中的随机排布。此问题是该堆型设计中面临的主要挑战之一。本文基于MCNP程序和固态燃料钍基熔盐堆（TMSR-SF1）模型完成了不同燃料球床与燃料球描述对关键中子学参数（如k_eff、堆芯能谱、控制棒价值和温度系数等）的影响分析。燃料球床描述使用随机序列添加（RSA）方法建立了随机球床模型与体心立方（BCC）结构的等效规则模型。包覆燃料颗粒描述则基于简立方（SC）等效模型利用MCNP程序中的URAN卡实现随机扰动。结果表明,包覆燃料颗粒随机分布的影响远小于燃料球随机分布的影响;尽管具有相同的总堆积密度,等效规则模型相比于随机球床模型会增加堆芯中子的泄漏,低估冷态满装载反应性约0.5%,高估控制棒价值约5%。     

熔盐/辐照环境中细结构等静压石墨的结构和化学稳定性研究

熔盐堆以石墨作为反射体和慢化体,熔盐与石墨直接接触,石墨在熔盐中的腐蚀反应和辐照损伤是值得研究的问题。本文采用自主研发的细结构石墨,阻隔熔盐浸渗,采用30 MeV He+模拟中子辐照,研究不同温度及熔盐环境下石墨微观形貌、微结构和化学结构的变化。研究结果表明,高温环境下,由于高温的退火效应,石墨缺陷密度的增加及形貌的变化都远小于室温环境。辐照后的石墨与熔盐接触,其缺陷密度略微降低。这种微结构的改善与高温熔盐环境中的退火效应及熔盐固化引起内部微裂纹的闭合有关。辐照后的熔盐浸泡可在石墨C—C键结构中引入C—F键,且C—F键的形成与缺陷密度及缺陷类型密切相关。稳定的空位簇及间隙原子的迁移均会影响层间化合物的形成,从而产生限制C—F键形成的环境,进而降低由层间化合物的形成对石墨表面结构的破坏。     

Protection of nuclear graphite toward fluoride molten salt by glassy carbon deposit

Molten salt reactor represents one of the promising future Generation IV nuclear reactors families where the fuel, a liquid molten fluoride salt, is circulating through the graphite reactor core. The interactions between nuclear graphite and fluoride molten salt and also the graphite surface protection were investigated in this paper by powder X-ray diffraction, micro-Raman spectroscopy and scanning electron microscopy coupled with X-ray microanalysis. Nuclear graphite discs were covered by two kinds of protection deposit: a glassy carbon coating and a double coating of pyrolitic carbon/glassy carbon. Different behaviours have been highlighted according to the presence and the nature of the coated protection film. Intercalation of molten salt between the graphite layers did not occur. Nevertheless the molten salt adhered more or less to the surface of the graphite disc, filled more or less the graphite surface porosity and perturbed more or less the graphite stacking order at the disc surface. The behaviour of unprotected graphite was far to be satisfactory after two days of immersion of graphite in molten salt at 500 °C. The best protection of the graphite disc surface, with the maximum of inertness towards molten salt, has been obtained with the double coating of pyrolitic carbon/glassy carbon.     

Gas absorption and discharge behaviors of lead-lithium

The absorption of rare gas in the lead-lithium has been quite low and the gas is used as a cover-gas to control the environment of experiment. In our previous thermo-fluid experiment by using lithium-lead, it was found the cover gas pressure enclosed in the very leak tight container of lithium-lead was decreased with time, that is, the gas-absorption of the solid lithium-lead occurred at room temperature under atmospheric pressure. The variation of pressure exceeded the retention of argon in lead-lithium which is expected by the published data. Therefore, we aim to confirm those phenomena under well-controlled experimental condition by using argon, nitrogen and helium. According to the results of gas exposure tests, the absorption of argon in the lead-lithium is comparable with that of helium even at the solid state. For the molten state of lead-lithium, the absorption of argon could be larger than that of helium. It is also observed from the TPD-MS analysis that the argon tends to desorb when the phase change of lead-lithium occurs. If the retention of argon in the lead-lithium cannot be ignored, the problem of Ar-41 activity should be taken into consideration as well as the problem of argon bubble in the lead-lithium.     

熔盐冷冻壁应用中关键工艺影响因素研究

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,而冷冻壁厚度的稳定控制是干法处理流程应用冷冻壁技术实现处理工艺目的的关键。基于自行研制的冷冻壁实验装置,模拟了干法处理中熔盐冷冻壁的应用工况,考察了导热油进口温度、熔盐初始温度、加热器功率、冷冻壁初始厚度对冷冻壁厚度变化的影响,得到了各个因素的影响规律,并总结了最佳的应用工艺条件。利用热流量的变化分析了冷冻壁厚度变化的原因:热流量越大,冷冻壁厚度减小量越大,达到平衡时,热流量越大,冷冻壁平衡厚度越小。通过实验数据拟合得到了线热流密度与冷冻壁平衡厚度的关系式,平均相对误差11.2%。     

高温熔盐试验回路系统设计及验证研究

The Thorium Molten Salt Reactor (TMSR) project plans to construct a 2 MWt liquid fuel molten salt reactor. After successful R&D of the proto-type equipment like pump, heat exchanger and freeze valve, construction of a high temperature fluoride loop is designed and constructed to test them. The rated power of heat exchange is designed to be 200 kW, and the flow rate is set as 15 m3/h by the caloric equation. Hence the pipe caliber chose to be DN50. The numerical computation by Fluent revealed that the pressure drop of the loop system is about 155 kPa, and the overdesign of the pump head is set as 20 m accordingly. To avoid thermal stress concentration of the pipeline, both the electric heater and radiator are designed to be fixed on a flexible universal sphere support, except the pump which is anchored on steel support directly. The structural design of the equipment and the loop hase been verified by several tests with an accumulated operating time near 4000 hours. The actual pressure drop of the loop is about 110~120 kPa, which still needs to be accurately measured by differential manometer in future. The analysis of the impurity of the molten salt shows that the contents of Cr and Mo increased by two orders of magnitude after loop operation, which means that an obvious material corrosion occurs. The level of moisture and oxygen, which is the main reason that cause fluoride corrosion, should be controlled considerably lower than the original design level of 100 μL/L.