Development of Electrochemical Separation Methods in Molten LiF-NaF-KF for the Molten Salt Reactor Fuel Cycle

Electrochemical methods for the separation of fission products from fission material in molten fluoride salt media have been studied in the context of their application within the framework of the developed Molten Salt Reactor fuel cycle. The separation possibilities of selected actinides (U, Th) and lanthanides (Nd, Eu, Gd) in molten eutectic LiF-NaF-KF at 530°C were evaluated by means of cyclic voltammetry. The applicability of different electrochemical techniques is discussed with reference to the new results from this study, and a basic flow sheet for the Molten Salt Reactor fuel cycle is outlined.     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

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。     

液态熔盐堆堆芯功率模糊PID控制

液态熔盐堆采用熔融氟化盐为燃料，燃料熔盐出口温度是衡量熔盐堆安全的重要指标。通过堆芯功率控制可实现燃料熔盐出口温度控制。将液态熔盐堆堆芯划分成内区和外区，并基于能量守恒原理建立堆芯非线性模型，采用微扰理论对非线性模型进行线性化。基于堆芯线性化模型，采用模糊比例-积分-微分（PID）控制器设计堆芯功率控制系统。以熔盐增殖堆（MSBR）为例，开展堆芯功率控制仿真。结果表明，引入10-3、2×10-3阶跃反应性时，模糊PID控制器可以减小系统响应的上冲幅度和超调量，并且在堆芯功率发生了较大的负荷变化时，模糊PID控制器可以对堆芯功率的变化实现良好跟踪。故所采用的模糊PID控制器具有良好的动态性能，可实现对堆芯功率的良好控制。      

Investigation of the MSFR core physics and fuel cycle characteristics

The adoption of Th fuel in fast reactors is being reconsidered due to the potential favorable impact on actinide waste management and resource availability. A closed Th cycle leads to an actinide inventory with lower radiotoxicity and heat load for the first several thousands of years. Due to the typically low TRansUranic (TRU) Conversion Ratio (CR), Th can also be advantageous to expedite the consumption of legacy TRU. One of the main obstacles to the implementation of Th is the highly radioactive recycled fuel which requires remote handling under heavy shielding, inevitably penalizing economics and challenging conventional pin-based fuel manufacturing. From this perspective, the development of liquid-fuelled reactors, with Molten Salt Reactors regarded as the most promising, appears particularly attractive as fuel handling would be greatly simplified. The present paper investigates the fuel cycle performances of the reference GEN-IV Molten Salt Fast Reactor (MSFR) in terms of isotope evolution, radiotoxicity generation and safety-related parameters. Similarly to most MSR concepts proposed in the past, the MSFR is based on the fluoride molten salt technology, but it features the novelty of a fast neutron spectrum. Calculations are performed using state-of-the-art equilibrium-cycle methodologies, i.e., the ERANOS-based EQL3D procedure developed at the Paul Scherrer Institut and extended to the simulation of the MSFR. Selected results have been benchmarked with the Monte Carlo code PSG2/SERPENT. These results have also been used for the assessment of a diffusion module based on the COMSOL multi-physics toolkit, which is the subject of current studies aimed at efficiently simulating the peculiar MSFR transient behavior.     

Molten salts and nuclear energy production

Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed.     

熔盐冷却球床实验堆堆内热源分布计算与分析

熔盐冷却球床堆采用球形燃料元件，冷却剂采用高温熔盐，其堆内热源分布与压水堆有着明显的区别，而与同样使用球形燃料元件的高温气冷堆相比，燃料球产生的中子和γ会在冷却剂中沉积更多的能量，因此准确计算堆内释热率分布对于这种新型反应堆的热工水力设计、瞬态分析、结构力学设计等都有重要意义。本文使用蒙特卡罗计算程序MCNP对中国科学院设计的10 MW固态燃料钍基熔盐实验堆（TMSR-SF1）堆内的释热率分布进行了详细计算研究，通过使用光子产生偏倚卡（pikmt），经过3次MCNP输运计算得到了TMSR-SF1寿期初（BOL）及寿期末（EOL）堆内各部件的总释热率、体积释热率分布和最大体积释热率。计算结果显示，燃料球释热率占堆内总释热率的94%以上，熔盐和反射层释热率占总释热率的1%以上，其他堆内部件释热率的比例都小于1%。寿期末燃料球、控制棒与石墨球的释热率均有所减少，而反射层等其他构件的释热率有所增加。     

Development of hybrid reprocessing technology based on solvent extraction and pyro-chemical electrolysis

Toshiba has been proposing a new fuel cycle concept for the transition period from Light Water Reactors (LWRs) to Fast Reactors (FRs). This concept involves a more valuable process for LWR spent-fuel reprocessing than the conventional process and improved proliferation resistance. We have been developing a new technology, the Toshiba Hybrid Reprocessing Process, based on solvent extraction and pyro-chemical electrolysis, for spent fuel reprocessing for the transition period from LWRs to FRs. The Toshiba Hybrid Reprocessing Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high-purity uranium (U) and the pyro-chemical process in molten salts for recovery of impure plutonium with minor actinides (Pu + MA). High-purity U is used for LWR fuel, and impure Pu + MA is used for metallic FR fuel. Valence control by the electrolysis and solvent extraction tests using LWR spent fuel and oxalate precipitation tests were carried out to confirm the feasibility of the Toshiba Hybrid Reprocessing Process. A consecutive processing equipment for the solvent extraction process and a bench-scale apparatus for the pyro-chemical process were manufactured. The consecutive processing equipment consists of a flow type electrolytic cell and a centrifugal extractor. The test revealed that U of 99.99% of purity was recovered. The bench-scale apparatus consists of a reactor for oxalate precipitation, a solid-liquid separator in which nitric acid with fission products and precipitation are separated, and a drying equipment in which the precipitation is dry. Precipitation test with neodymium (Nd) which is simulated as Pu + MA in nitric acid was carried out. It was confirmed that precipitation ratio of Nd was more than 99.9% and that moisture ratio of the precipitation was less than 10%. The results suggested that U recovery of LWR spent fuel was 99.99% with the consecutive processing equipment and Pu + MA recovery was more than 99.9% with the bench-scale apparatus. The Toshiba Hybrid Reprocessing Process could recover high-purity U used for LWR fuel, and impure Pu + MA used for metallic FR fuel.     

钍基熔盐堆燃料循环与启动策略研究

研究了熔盐燃料在堆内外循环以及考虑特殊核素的添加、提取等在线处理过程的熔盐堆燃耗计算模型，在多功能组件计算程序SONG的基础上开发了相应的燃料循环计算功能并进行了初步验证。在此基础上，分别针对氧化铍慢化的热谱熔盐堆和无慢化的快谱熔盐堆进行计算，并根据堆芯反应性长期稳定的基本要求，分析了利用²³³U和工业Pu启动熔盐堆时配套的在线处理方案以及相应的易裂变核添加要求。通过对核素添加、提取以及燃料内核密度的平衡计算，分析了不同的在线处理方案与启动策略对钍-铀燃料循环效率的影响，并据此提出了初步的熔盐堆燃料循环技术路线。结果表明：压水堆乏燃料提取的工业Pu较²³³U更适宜用于钍铀燃料循环启动，因工业Pu启动的快谱熔盐堆的²³³U产率明显高于²³³U启动熔盐堆，而当有了足够的²³³U积累后，²³³U启动的热谱熔盐堆是更好的选择，因其燃料倍增时间更短且燃料初装量也小得多。     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

基于SCALE的熔盐堆添料与后处理系统分析程序开发及验证

在线添料及在线去除中子毒物是熔盐堆区别于其他固体燃料反应堆的主要特征之一，能够实现较高的燃耗深度和燃料利用率。然而，现有的反应堆物理计算分析软件SCALE不能直接模拟熔盐堆的燃耗计算。因此，本文耦合SCALE中的截面处理模块、临界计算模块以及燃耗计算模块，开发了一套适用于多流体熔盐堆的添料与后处理系统分析程序MSR-RRS，实现熔盐堆的在线添料、裂变产物在线处理或离线批次处理等模拟功能。基于MSR-RRS对现有的单流熔盐增殖堆和双流熔盐快堆的燃耗性能进行了验证。结果表明，MSR-RRS计算结果与基准模型结果符合较好。MSR-RRS适用于多种堆型、多种燃料循环运行模式。     

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%。     

乏燃料后处理湿法工艺技术基础研究发展现状

为了保持核能可持续发展,必须相应发展乏燃料后处理技术,以实施快堆闭合核燃料循环。湿法后处理工艺仍以PUREX流程为基础,从乏燃料元件首端处理工艺、萃取工艺的简化和无盐调价等方面开展相应的研究。同时随着动力堆乏燃料元件燃耗的增加,Np、Pu以及高产额裂变产物元素Ru、Tc、Zr等在水法后处理工艺中的行为及形态等影响日趋凸显。本文针对上述问题进行了论述,并提出了相应的研究重点。     

基于MOREL2.0对石墨慢化熔盐堆的稳态分析

熔盐堆采用熔融的氟化盐混合物作为燃料和堆芯的冷却剂，由于燃料的流动，熔盐堆在中子学和热工水力学方面与传统固体燃料反应堆有着较大区别。本文基于熔盐堆分析程序MOREL2.0对钍基熔盐堆（TMSR）初步堆芯设计方案进行了稳态计算分析，结果表明：燃料流动对缓发中子先驱核的分布影响较大，并导致169 pcm反应性损失；随燃料在外部回路中滞留时间的增加，k_eff降低，80 s后趋于平稳；TMSR具有负的入口燃料温度系数，具有固有安全性。     

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。     

基于空气冷却的熔盐堆非能动余热排出系统优化设计

上海应用物理研究所基于TRISO包覆球形颗粒燃料与液态氟盐提出了基于钍基熔盐固态试验堆（TMSR-SF1）技术方案,其中一个重要的工作是非能动余热排出系统（PRHRS）设计。由于熔盐与水的不兼容特性,以及其高运行温度,采用空气作为最终热阱来设计PRHRS成为必然。为实现系统最简化、体积最小化以及排热与保温兼顾的设计目标,本文从MSR堆芯活性区到外界空气热阱传热过程的模型入手,建立了PRHRS优化设计模型,获得了优化设计方案,并基于改进的RELAP5/MOD4.0程序（针对TMSR-SF1的专门改进程序）开展了PRHRS容量论证评价,经计算分析,PRHRS容量设计合理,可确保反应堆全厂断电（SBO）后排热安全。      

Prospective Power-Reactor Fuel Cycles Based on Water-Free Reprocessing of Spent Fuel

A pyroelectrochemical process for reprocessing spent fuel and fabricating granular oxides UO₂, PuO₂ or (U, Pu)O₂ from chloride melts has been developed at the Scientific-Research Institute of Nuclear Reactors for a prospective nuclear fuel cycle. The basic equipment has been developed. The basic results of a comprehensive study of fuel elements with vibrationally compacted (U, Pu)O₂ fuel for fast reactors are presented. The performance of the reactors remains high up to 30% burnup in standard BOR-60 reactor fuel assemblies and 32% burnup in experimental fuel elements. An assessment is made of the effectiveness of the pyroelectrochemical methods and vibrational compaction technology for plutonium utilization.     

The reactor dynamics code DYN3D – models,validation and applications

The article provides an overview of the reactor dynamics code DYN3D. The code comprises various 3D neutron kinetics solvers, a thermal-hydraulics reactor core model and a thermo-mechanical fuel rod model. The implemented models and methods and the capabilities and features of the code are described. Latest developments of models and methods are delineated. An overview on the status of verification and validation is given. Code applications for selected safety analyses are described. Furthermore, multi-physics code couplings to thermal-hydraulic system codes, CFD and sub-channel codes as well as to the fuel performance code TRANSURANUS are outlined. Developments for innovative reactor concepts, in particular Molten Salt Reactor, High Temperature Gas-cooled Reactor and Sodium Fast Reactor are delineated. The management of code maintenance is briefly described. An outlook on further code development is given.