钍基熔盐堆新型栅格设计与优化

利用上海核工院自主开发的SONG/TANG-MSR程序系统,通过大量的方案筛选,在熔盐燃料成分、慢化剂材料以及栅格结构等方面对钍基熔盐堆（TMSR）栅格进行设计优化。提出采用无铍熔盐作为燃料,以提高重核（ThF4, UF4）溶解度;采用氧化铍（BeO）作为慢化剂,从而提高中子经济性;创新性地采用熔盐燃料与慢化剂隔离的SiC包壳管设计,保持堆芯结构的稳定性和抗辐照性能。对优化后的栅格设计计算结果表明:新型的熔盐堆栅格设计具有很高的增殖比并保持负功率系数,从而满足下一代核能系统可持续性和安全性要求。     

新型钍基熔盐堆堆芯方案及燃耗分析

采用自主开发的SONG/TANG-MSR栅格/堆芯分析程序对新型钍基熔盐堆（TMSR）进行堆芯布置与燃耗分析计算。根据前期的栅格分析相关工作,TMSR采用了无铍（BeF2）燃料熔盐、氧化铍慢化剂以及碳化硅包壳,并在组件栅格初步优化分析的基础上,通过全堆芯计算对熔盐栅格进一步优化和分析,给出了堆芯三区布置方案。该方案具有较高的增殖比,负的功率系数,以及较平的温度分布。根据该堆芯方案,在考虑熔盐在线处理情况下进行了熔盐燃耗计算分析。结果表明,堆芯具有较高的增殖比、较短的倍增时间以及长期稳定运行能力。新型的钍基熔盐设计大大提高了增殖性能,同时又确保堆芯具有足够的安全性能。     

TMSR白光中子源关键核数据实验进展

钍铀燃料循环核数据的精度和可靠性直接关系着钍基熔盐堆的安全性和经济性。目前大多数核数据都是基于铀钚燃料循环进行开发,若直接用在钍基熔盐堆上将会出现核设计不确定度较高的问题。为了提高钍基熔盐堆物理设计所需核数据的适用性,中国科学院上海应用物理研究所自行设计并建造了紧凑型的15 MeV电子加速器驱动的白光中子源(Photoneutron Source,PNS),用于开展钍铀燃料循环核数据的实验测量。该装置已通过技术验收,并进行了一系列关键核素的核数据测量,检验了现用核数据的可靠性,为相关核素的核数据评价与改进提供了基础实验数据。     

钍基熔盐堆核能系统

裂变核能是保障能源增长需求和促进节能减排的重要手段。钍基核能具有中子增殖性能好、产生高放废料少和储量丰富等特点;熔盐堆作为第四代先进反应堆的6个候选堆之一,具有钍高效利用、高温制氢、无水冷却、适合小型模块化设计等优势和潜力。发展钍基熔盐堆,对中国的核能战略发展具有重要意义,将会有助于解决中国的能源和环境双重挑战。2011年1月,中国科学院启动战略性先导科技专项——钍基熔盐堆核能系统(TMSR),采取液态熔盐堆和固态熔盐堆两种堆型研发同时部署和相继发展的技术路线,致力于实现基于熔盐堆的钍资源高效利用和核能综合利用,为我国核能发展赢得先机。     

液体燃料熔盐堆三维瞬态耦合分析

根据熔盐堆的流动特性,采用多群扩散理论,建立液体燃料熔盐堆的三维中子动力学模型和流动传热模型,开发针对液体燃料熔盐堆的三维核热耦合程序,并分析瞬态情况下MOSART堆的物理热工特性.结果表明：堆芯热工参数的瞬态变化直接影响物理参数的变化,液体熔盐堆具有较明显的温度负反馈特性.     

钍铀循环关键核素的多群数据库加工与初步验证

钍基先进核能系统的发展对核数据提出了新的需求。钍铀循环关键核素的多群数据库是堆芯物理设计与分析计算的基础。因此,基于权威核数据处理程序NJOY对ENDF/B-VII.1评价核数据库进行处理得到WIMS格式多群截面数据,利用专用更新处理程序WILLIE得到WIMS格式数据库,并利用多群堆芯计算程序WIMSD-5B对数据库进行了一系列临界基准检验,结果表明:基于ENDF/B-VII.1加工的WIMS库与WIMS库更新计划（WLUP）的网站上发布的最新版本WIMS库的计算结果基本一致,并且,在16个钍铀循环基准题检验中,新加工的WIMS库计算平均误差要比WIMSD5B程序的自带WIMS库的计算结果小0.225 3%,精度更高,可靠性更好。     

新概念熔盐堆堆芯稳态热工水力计算

针对第四代国际核能论坛提出的新概念熔盐堆堆芯的具体特点,采用DRAGON程序计算了堆芯轴向和径向功率分布因子,并对熔盐堆的燃料盐和石墨慢化剂分别建立数学模型,通过数值方法对新概念熔盐堆的稳态热工水力特性进行了模拟.计算得到了堆芯燃料盐的流量分配,代表元件内燃料盐和内外壁温度沿轴向的变化.     

重入式谐振腔电磁场建立的瞬态物理过程

 采用理论推导、计算模拟和实验测量3种方法,对重入式谐振腔电磁场稳态建立的瞬态物理过程进行研究,得到的结果相一致。阐明了电磁场建立的瞬态物理过程,结果显示：电磁场稳态建立时间与谐振腔的固有谐振频率成反比,与品质因子成正比;仿真模拟和理论计算偏差小于0.8%,实验测试和理论计算偏差小于2.5%。     

小型模块化熔盐堆钍利用方案

钍基熔盐堆（Thorium Molten Salt Reactor,TMSR）核能系统先导专项的研究目标是研发第四代裂变反应堆核能系统（即钍基熔盐堆）。为充分利用液态燃料熔盐堆的在线添料与在线燃料处理的优势,同时考虑熔盐堆的快速部署,TMSR先导专项部署了小型模块化熔盐堆。考虑燃料处理技术现状及其可能的发展方向,小型模块化熔盐堆钍利用方案采用"三步走"战略。第一阶段采用在线加料与离线处理,实现钍的成规模利用;第二阶段采用在线加料和在线处理（U）与离线处理（MA）的结合,实现钍的高效利用;第三阶段采用在线加料及在线处理全部重金属,实现钍的自持增殖利用。随着"三步走"战略的逐步实施,钍铀燃料循环模式及后处理性能稳步提高,重金属利用率得到明显改善,同时有效降低了卸料毒性。考虑燃料许可容易度和建堆时间,首先为钍利用方案第一阶段布置了三种可能的启堆燃料,分别为低富集铀、低富集铀加钍和233U加钍。计算结果显示:以低富集铀启堆时,燃料循环性能与水堆相当;以233U启堆时,燃料利用率明显高于水堆,且其放射性毒性比水堆低约2个数量级。The missions of the Thorium Molten Salt Reactor (TMSR) Nuclear Energy System are to research and develop the thorium based molten salt reactors (MSR) belonging to the fourth generation of nuclear fission reactor system. A Small modular Molten Salt Reactor (SmMSR) is deployed to make full use of the advantages of online refueling and online reprocessing and to consider the rapid deployment of MSR. An innovative "three-stage" strategy of thorium utilization based on SmMSR is proposed to take the current condition of fuel reprocessing and its future evolution. The first stage can realize the thorium utilization at a large scale with online refueling and off-line processing. The second stage can obtain efficient thorium utilization with online refueling, online processing of uranium and off-line processing of minor actinides (MAs). The third stage is implemented with self-sustaining or breeding mode with online refueling and online processing of all heavy metals. Along with the development of three stages, the utilization of heavy metals will be obviously improved and the radio-toxicity will be significantly reduced. A SmMSR is designed to achieve the goals of the first stage of thorium utilization. And three kinds of nuclear fuel cycles with different startup fuel types (i.e., low enriched uranium (LEU), thorium mixed with LEU (LEU+Th) and thorium mixed with 233U (233U+Th)) are implemented. The results show that the performance for fuel cycle containing LEU is comparable to the pressurized-water reactor (PWR). Meanwhile, the nuclear utilization for that containing 233U is much higher than PWR, and the radio-toxicity for which is lower by ~2 magnitudes than that for PWR.     

TMSR白光中子源本底屏蔽设计

为了满足钍基熔盐堆对核数据的需求,中国科学院上海应用物理研究所自行设计并建造了紧凑型的15 MeV电子加速器驱动的白光中子源。电子直线加速器、中子产生靶以及探测器系统处于同一个实验大厅,中子/伽马射线本底较高,原有屏蔽并不能满足在低能区进行热中子物理实验测量的低中子本底需求。为了降低热中子本底,提高在热区的测量能力,需要对中子源进行局部屏蔽。根据调试运行经验以及模拟计算结果,分析了中子伽马射线本底的来源,利用MCNP5模拟计算了混凝土、铅、含硼聚乙烯对中子/伽马射线的屏蔽效果,优化设计了局部屏蔽方案。模拟计算结果显示,该屏蔽方案可将热中子本底降低三个量级,伽马本底降低两个量级。屏蔽后的实验测量结果表明,探测器处的有效热中子与本底热中子的比值达到约100:1,屏蔽效果显著,为后续在热中子能区顺利开展中子物理测量实验奠定了基础。     

Methods and computer codes for nuclear systems calculations

Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space-time calculations. Computer code TRIFON solves space-energy problem in (X, Y) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space-time neutron processes simulation. Modification of TRIFON was developed for the simulation of space-time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.      

Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.     

基于FLUENT的核热耦合程序反应性反馈参数敏感性

将计算流体力学模型与中子动力学模型耦合来进行反应堆瞬态安全分析的方法,由于可以开展复杂几何结构的三维流动传热分析,因此受到很大的关注。基于FLUENT用户自定义功能（UDF）开发了一套可用于池式铅堆瞬态安全分析的核热耦合程序,程序耦合了临界/次临界点堆中子动力学模型和燃料棒模型。由于反应堆处于不同寿期时,随着燃料燃耗、可燃毒物积累等因素导致反应性反馈系数有较大变化,因此使用开发的核热耦合程序对中国科学技术大学提出的小型自然循环铅冷快堆进行不同关键反馈系数下无保护的瞬态超功率事故安全分析。调整点堆模块考虑到的四个反应性反馈系数,可以发现燃料多普勒系数对堆安全的影响最大,同时定量的分析结果表明超功率事故引入时间长短对事故演化有重要影响。     

Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.     

新型低熔点熔盐黏度的实验研究

熔盐因其具有广泛的使用温度范围,低蒸气压,大热容量,低黏度,良好的稳定性,低成本等诸多特性已成为聚光太阳能热发电中颇有潜力的传热蓄热介质。准确的熔盐热物性对于太阳能发电过程中介质的传热蓄热性能有重要影响。其中熔盐黏度作为重要的热物性之一,对于提高传热效率和降低流动阻力具有决定作用。本文利用研制的高温黏度测量仪对水和HITEC盐的黏度温度特性进行了实验研究,实验结果与文献数据具有较好的一致性,证明了该高温熔盐黏度仪的可靠性。为了降低混合熔盐的熔点,改进其热物性能,本文对Solar Salt进行改性研究,得到两种新型低熔点混合熔盐,并测定得到了黏度温度特性曲线。结果表明,改性后的高温熔融盐黏度有所降低,有利于降低太阳能热发电熔盐传热管路系统的阻力和成本。