基于熔盐快堆的模型优化与Th-U增殖性能研究

作为四代堆6种候选堆型中唯一的液态燃料反应堆,熔盐堆对未来核能和钍资源利用具有重要意义,特别是熔盐快堆(Molten Salt Fast Reactor,MSFR)还具有较大的增殖比和较好的温度负反馈。由于启动新的熔盐快堆需要较高的燃料装载量,若能改善MSFR的增殖性能,则有利于提高233U产量并缩短燃料倍增时间。首先应用SCALE6.1针对MSFR的径向增殖盐、新增轴向增殖盐和新增石墨反射层这三方面分析了初始增殖比,同时从核素吸收率角度说明增殖比变化的原因和MSFR的设计不足并对其进行了优化;然后应用基于SCALE6.1开发的熔盐堆在线处理模块(Molten Salt Reactor Reprocessing Sequence,MSR-RS)进行燃耗分析。结果表明,新增轴向增殖盐可以进一步提高增殖性能;新增石墨反射层可以节省增殖盐装载量。改进后的堆型运行时增殖比可以维持在1.1以上,233U年产量提高至133 kg,倍增时间缩短至36 a,并且堆芯在整个运行寿期都能保持足够的温度负反馈。     

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。     

小型模块化熔盐快堆燃料管理初步分析

由于燃料随熔盐流动的特性以及可以进行在线添料与处理的特点,液态燃料熔盐堆的燃耗分析与燃料管理和传统固态燃料反应堆有很大不同,需要针对液态燃料熔盐堆的特点重新开发燃耗分析与管理程序。本文针对液态燃料熔盐堆的熔盐流动特性以及在线添料与处理功能,基于MCNP5和ORIGEN2.1燃耗耦合程序,开发了适用于液态燃料熔盐堆的燃料管理程序,并应用于一种小型模块化熔盐快堆的燃料管理和分析,对比分析了5种不同运行方案以及分批在线添料情况下,运行30年期间keff的变化情况及重要核素的演化情况。计算结果表明,采用不断调整添料率的连续在线添料运行方案和固定批量添料的运行方案,都可以让小型模块化熔盐快堆维持运行在一个较小的keff波动范围之内。开发的燃料管理程序适用于液态燃料熔盐堆的研究,同时可以为液态燃料熔盐堆的设计及燃耗管理和分析提供有价值的参考。     

熔盐燃料对加速器驱动的次临界堆中子学性能的影响

加速器驱动的次临界熔盐堆(Accelerator-Driven Subcritical Molten Salt Reactor,ADS-MSR)结合了熔盐堆与ADS的许多优点,在先进核燃料利用方面有独特的优势。为了研究熔盐燃料的使用对ADS系统堆芯的中子学性能的影响,基于MCNP(Monte Carlo N Particle Transport Code)程序,分别计算并分析了熔盐燃料对加速器驱动的次临界堆的外源质子效率、中子能谱以及钍铀转换比等参数的影响。结果表明:相较于氧化物燃料,熔盐燃料的使用将会增加对外源中子和裂变中子的慢化,并且会提高堆芯的入射质子效率。同时,由于熔盐燃料的慢化效应,FLi Be和FLi熔盐燃料燃耗初期的钍铀转换比(CR)分别为1.023和1.062,略低于氧化物燃料的1.068。另一方面,熔盐燃料的在线处理会极大降低燃耗过程中的反应性损失。通过在线燃料处理和在线添料,FLi熔盐和FLi Be熔盐燃料的CR分别在燃耗运行的第1年和第3年超过氧化物燃料,并且能够长期稳定在1.06和1.00左右。     

熔盐堆燃料管理程序开发

由于燃料盐的流动,熔盐堆的燃料将在整个堆芯内重新分配,同时燃料中的裂变产物通过在线处理装置进行提取,这就使得熔盐堆内燃料成分变化更加剧烈,传统的固体燃料反应堆的宏观燃耗管理程序已不再适用。本文采用基于解析基函数展开法的扩散理论及线性子链微观燃耗算法开发了基于"两步法"的微观燃耗管理程序MOREL,并进行了验证。结果表明,MOREL程序有很好的精度,最后对钍基熔盐堆进行了初步的换料分析。     

熔盐堆中燃料流动对缓发中子的影响分析

熔盐堆具有良好的中子经济性、固有安全性、可在线后处理、防核不扩散等特点,是六种第四代先进反应堆堆型中唯一的液体燃料反应堆。然而,熔盐堆中采用流动的熔盐作为液体燃料,从而缓发中子先驱核会随着燃料的流动流出堆芯并在堆芯外发生衰变,这不同于固体燃料反应堆。文中针对了一座实际运行过的熔盐实验堆(Molten Salt Reactor Experiment,MSRE),基于中子动力学模型,采用圆柱体均匀堆的近似处理方法推导了液体燃料反应堆的缓发中子先驱核浓度数学模型,研究了恒定流速下的反应性损失及不同燃料熔盐流速对缓发中子分布的影响。结果表明缓发中子在越靠近堆芯中心区域的位置就越多,同时熔盐流速的变化对衰变周期越短的缓发中子先驱核组数的影响比较小。通过本研究,可以了解熔盐堆中缓发中子随着燃料流动的变化情况,为熔盐堆安全分析提供参考依据。     

熔盐堆物理热工耦合程序开发及验证分析

熔盐堆(Molten Salt Reactor,MSR)是第四代反应堆6种堆型中唯一的液态燃料反应堆,与固态燃料-液体冷却剂反应堆相比,原理上有较大不同。在熔盐堆中,流动的熔盐既是燃料又是冷却剂与慢化剂,中子物理学与热工水力学相互耦合;由于熔盐的流动性,缓发中子先驱核会随燃料流至堆芯外衰变,造成缓发中子的丢失,导致堆芯反应性降低。正是由于熔盐堆的这些新特性,造成熔盐堆内缓发中子先驱核、温度等参数变化与固态燃料反应堆有所不同,需要研究熔盐堆在各种工况下的相关物理参数变化。本文主要工作是考虑缓发中子先驱核的流动性对熔盐堆的影响,研究适用于熔盐堆的二维圆柱几何时空中子动力学程序及与之耦合的热工水力学程序;利用该程序对熔盐堆中子物理学和热工水力学进行耦合计算,验证熔盐堆相关实验数据;并且计算了熔盐堆无保护启停泵及堆芯入口温度过冷过热工况,用于分析熔盐堆的安全特性。计算结果表明,程序能够对熔盐反应堆实验(Molten Salt Reactor Experiment,MSRE)的相关实验数据进行较好的模拟计算,并且验证了熔盐堆的固有安全性。     

双区氯盐快堆的增殖及嬗变性能分析

基于熔盐嬗变堆(Molten Salt Actinide Recycler and Transmuter,简称MOSART)堆芯结构对氯盐快堆(Molten Chloride Salt Fast Reactor,简称MCFR)进行了优化,分析了熔盐成分和后处理方式的影响,使其燃耗性能得到明显的提升,但是相比熔盐快堆(Molten Salt Fast Reactor,简称MSFR)的增殖及嬗变性能仍有一定差距。基于在线连续添料与后处理方式,采用SCALE6.1程序和熔盐堆在线添料和后处理程序(Molten Salt Reactor Reprocessing Sequence,简称MSR-RS)分析了堆芯结构、~(37)Cl富集度对增殖比(Breeding Ratio,简称BR)、核素吸收率、燃耗等方面的影响,提出了双区氯盐快堆的设计,进一步提升了增殖嬗变性能和钍基燃料的利用率,倍增时间缩短到20年左右,超铀核素(Transuranics,简称TRU)嬗变率达到68%左右。     

液态燃料熔盐堆三维动力学程序开发及验证

液态燃料熔盐堆的燃料熔盐在一回路中循环流动,一回路高温熔盐既是燃料,又是冷却剂,大部分核裂变能直接释放在燃料熔盐之中。随着燃料熔盐流动,一部分缓发中子先驱核（Delayed Neutron Precursors,DNP）在堆芯外一回路中衰变引起反应性损失。液态燃料熔盐堆中子物理与热工流体紧密耦合,传统固态燃料反应堆堆芯核热耦合程序不再适用于液态燃料熔盐堆。针对液态燃料熔盐堆特点,建立了包含带对流项的DNP输运方程和带热内热源热工流体方程的液态燃料熔盐堆动力学模型,并基于节块展开法,开发了堆芯三维动力学程序ThorCORE3D。使用美国橡树岭国家实验室建造运行的熔盐实验堆（Molten Salt Reactor Experiment,MSRE）稳态和瞬态实验基准题,对ThorCORE3D程序进行了初步验证。结果表明：ThorCORE3D程序计算值与MSRE实验值吻合良好,适用于液态燃料熔盐堆稳态设计与瞬态分析。     

FLi/FLiBe盐中7Li富集度对熔盐快堆钍铀转换性能的影响研究

锂(Li)元素是液态熔盐堆中冷却剂熔盐的重要组成成分,由于6Li相对~7Li具有较大的中子吸收截面,其在冷却剂熔盐中的摩尔含量会影响液态熔盐堆的钍铀转换性能,因此研究~7Li富集度对液态熔盐堆钍铀转换性能的影响十分必要。基于熔盐快堆(Molten Salt Fast Reactor,MSFR)的堆芯结构,分别采用FLi和FLiBe两种不同的冷却剂熔盐,选取范围在99.5%～99.995%的一系列~7Li富集度,借助熔盐堆后处理程序MSR-RS(Molten Salt Reactor Reprocessing Sequence),针对能谱、233U初装量、钍铀转换比、233U净产量和倍增时间、Li的演化以及氚产量等一系列参数进行分析。研究结果表明:在MSFR的堆芯中,较FLiBe而言,采用FLi作载体盐能够获得更好的钍铀转换性能;当~7Li富集度由99.995%变为99.9%时,堆芯钍铀转换比降低约1.6%,氚产量增加约8%。综合考虑燃料制造成本和钍铀转换性能等因素,对于分别采用FLi和FLiBe作载体盐的熔盐快堆MSFR,推荐的~7Li富集度都为99.9%。     

Burnup optimization of once-through molten salt reactors using enriched uranium and thorium

The advantages of once-through molten salt reactors include readily available fuel,low nuclear proliferation risk,and low technical difficulty.It is potentially the most easily commercialized fuel cycle mode for molten salt reactors.However,there are some problems in the parameter selection of once-through molten salt reactors,and the relevant burnup optimization work requires further analysis.This study examined once-through graphitemoderated molten salt reactor using enriched uranium and thori...     

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.     

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.     

液态熔盐热堆燃料管理方法研究与分析

为研究液态熔盐热堆的燃料管理性能,需解决复杂堆芯结构的均匀化、燃料的混合及在线后处理3个问题。本文基于确定论程序DRAGON5与DONJON5,开发了液态熔盐热堆的燃料管理程序LMSR,并进行了验证。使用LMSR对液态熔盐热堆进行计算与分析,结果显示使用²³⁵U与²³⁸U启堆,加入燃料为²³²Th与²³³U条件下,后处理提取重金属的效率至少需要90%。此外,为维持堆芯有效增殖因数在1.0～1.005之间,加入的燃料中²³³U平均等效质量富集度在40%附近。     

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation’s energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.     

钠冷行波堆和驻波堆堆芯特性对比分析

采用自开发的MCNP-ORIGEN耦合程序MCORE对所设计的钠冷行波堆和驻波堆开展中子学和燃耗分析;基于MCORE获得的功率分布,采用自开发的钠冷快堆堆芯稳态热工水力分析程序SAST对钠冷行波堆和驻波堆堆芯开展热工水力分析。对比钠冷行波堆和驻波堆的堆芯物理特性和热工水力特性,结果表明：驻波堆在燃耗、最高包壳和燃料芯块温度方面具有优势,而行波堆在反应性波动和堆芯冷却剂出口温度均匀性方面具有优势。     

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.