2 MWt液态燃料钍基熔盐实验堆滞留盘管优化设计

液态燃料钍基熔盐实验堆滞留盘管是熔盐堆尾气处理系统的重要核安全级设备,为尾气中短寿命核素衰变提供封闭环境,同时去除衰变热,降低其后碘吸附床和活性炭吸附床的温度。本文介绍了业主提供的2 MWt液态燃料钍基熔盐实验堆滞留盘管原设计方案中存在的问题及结构优化后的新设计方案。优化后的滞留盘管设计方案具有散热性好、泄漏点少（焊缝少）、容易施焊、便于检测监控维修、用材少等优点,具有较好的可靠性和经济性。      

2MW液态钍基熔盐实验堆气载放射性流出物近场扩散的数值模拟

气载放射性流出物在近场范围内的扩散是核设施环境影响评价研究的重要内容之一,传统的高斯模型由于受到复杂建筑物的影响导致计算结果偏差比较大,不宜用于近场扩散的数值模拟。本文采用计算流体动力学(Computational Fluid Dynamics,CFD)方法以2 MW液态钍基熔盐实验堆的拟定场址为研究对象,开展放射性气态流出物在近场范围内分布规律的研究,分析风速、烟囱高度、风向等参数对气态流出物大气弥散因子分布的影响。结果表明,对于高架排放,由烟羽抬升的影响使得风速越大近场范围的放射性核素大气弥散因子越高;在下风向建筑群迎风侧均易出现放射性核素集聚区,烟囱高度越低集聚现象越明显。本研究的结果可为熔盐堆场区辐射环境影响评价及建筑物的布局、核应急提供参考依据。     

应用中子活化分析技术测定铀和钍放射性同位素的α能谱内标法

开发了一种用~(238)U和~(232)Th作为内标的α能谱测定环境样品中铀和钍放射性同位素的新方法。应用中子活化分析技术精确地测定样品中的~(238)U和~(232)Th。将其他部分样品全部溶解,采用阴离子交换法分离铀和钍,制备为α能谱测定的电镀薄(低强度)源。从已知的~(238)U和~(232)Th质量浓度,求得它们的活度浓度,之后据α能谱,从U(~(234)U、~(236)U、~(238)U)和Th(~(228)Th、~(230)Th、~(232)Th)放射性同位素的比值中,获得同位素的绝对活度浓度。该方法的优点是对放射化学分离的化学回收率和α能谱仪的计数率都没有特殊要求。应用内标可以不需要外加昂贵的、经校准的放射性同位素指示剂(如~(232)U和~(229)Th)。用一些已检定的环境参考样试验了这种新方法,并且同外加放射性同位素指示剂的传统方法进行了比较。其优点是同标准方法相结合可获得一组独立的数据,用于质量控制。     

熔盐快堆Th-U燃料循环增殖性能分析

熔盐快堆是当前国际上关注的热点之一,本文基于堆芯结构双流体方案,即裂变熔盐燃料和增殖熔盐介质各自独立冷却循环,利用氟化或氯化熔盐中钍铀重金属盐高温下的高溶解度特性,获得熔盐快堆的高增殖。通过比较钍铀燃料循环熔盐快堆的三种可行性熔盐燃料方案(LiF+ThF_4+UF_4、NaF+ThF_4+UF_4和NaCl+ThCl_3+UCl_3),采用基于反应堆安全分析和设计的综合性模拟程序SCALE(Standardized Computer Analyses for Licensing Evaluation),计算了中子能谱、反应性温度系数,分析了增殖比BR(breeding ratio)受反应堆裂变区、增殖区和ZrC中子反射层的尺寸影响、熔盐中~6Li和~(35)Cl同位素丰度的影响,以及熔盐密度误差对BR计算值的准确性影响、易裂变核素随反应堆运行时间演化等。在钍铀燃料循环熔盐快堆中,通过优化处理得到三种熔盐燃料方案的增殖比BR约为1.2。     

2MW液态钍基熔盐实验堆主屏蔽温度场分析

反应堆主屏蔽是核反应堆的重要组成部分,用来有效降低反应堆运行时屏蔽体外的辐射剂量水平,以满足反应堆部件材料对辐射限制的要求。温度是影响反应堆主屏蔽性能的重要因素。针对2 MWth液态熔盐堆(2-MW liquid-fueled molten salt experimental reactor,TMSR-LF1),采用MCNP软件获得功率分布后,利用Fluent软件对主屏蔽进行温度场计算。计算过程中利用Python语言编写了程序(MCNP to Fluent,MTF)来实现将MCNP(Monte Carlo N Particle Transport Code)计算结果转换为功率密度的空间分布,以用户自定义函数(User-Defined Function,UDF)形式导入到Fluent,解决了MCNP计算结果不能直接导入到Fluent的问题,并分别计算了TMSR-LF1熔盐堆不同环境温度下的主屏蔽温度场分布情况。结果表明,在环境温度为5°C、18°C、25°C、30°C、35°C、40°C情况下,TMSR-LF1熔盐堆主屏蔽普通混凝土墙温度均低于要求限值,达到设计要求。     

钍资源的核能利用问题探讨

分析了钍/铀燃料循环特点,评估了国际上钍资源利用研究开发现状和发展趋势,并试图按照科学发展观提出了我国钍资源核能利用的战略思考和钍/铀燃料循环前瞻性研究开发课题.     

液态燃料钍基熔盐实验堆主体装置厂房总体布置研究

为实现2 MW液态燃料钍基熔盐实验堆（TMSR）主体装置厂房的合理紧凑型总体布置设计，本研究根据熔盐堆堆型特征、顶层设计和系统功能需求，确定了主体装置厂房总体设计特征，探讨了TMSR关键设备及物项的相对位置特点；同时通过合理规划厂房功能分区和设备布置，最终得到了该厂房的总体布置方案。通过本项目的实施，为实现TMSR的系统集成以及验证提供了基础平台，为小型模块化钍基熔盐示范堆的设计和建设提供技术支持及经验。     

使用均匀混合型燃料组件的压水堆钍-铀增殖循环研究

为提升压水堆燃料利用率,设计了一种包含适量²³²Th和²³³U的均匀混合型燃料组件。对该型燃料组件的核特性分析表明,其具备随燃耗增加k_inf下降更缓慢的特性,有利于堆芯获得更长的循环长度。以岭澳核电厂一号机组为例,对包含均匀混合型含钍燃料组件的堆芯进行了分析,结果表明,当前压水堆中采用均匀混合型含钍燃料组件是可行的,并且具备²³⁵U利用率高、堆芯循环长度长的优势。     

重水堆钍铀燃料增殖循环方案研究

论文的目的是研究重水堆钍铀燃料增殖循环方案。基于前期设计的技术路线,以CANDU-6堆芯为参考堆芯,研究了钍基堆芯燃料管理策略,分析了中子学特性,并对乏燃料特性进行了评估,包括放射性毒性、衰变热和伽马射线。在此基础上,建立了钍铀燃料增殖循环方案,其在可持续性关键指标方面优于常规天然铀一次通过循环。     

基于热堆的钍铀转换过程中232U生成的模拟计算

在钍铀燃料循环过程中生成的232U的衰变子体具有强放射性,对燃料循环具有重要影响。本工作采用ORIGEN2、SCALE5程序,以及基于Bateman方法编写的程序,分析了在不同条件下,热堆中钍反应生成232U的规律。一般情况下,232U主要由232Th的(n,2n)反应链生成,而在中子能谱更软情况下,230Th对232U生成贡献增大;CANDU型重水堆和压水堆的含钍燃料组件的燃耗计算结果表明,铀中232U含量随燃耗深度增加而变大,同时初始230Th/Thtotal大小直接线性影响卸料燃耗时232U/Utotal或232U/233U。     

Radiation dose distribution of liquid fueled thorium molten salt reactor

A liquid fueled thorium molten salt reactor(TMSR-LF),one of the Generation IV reactors,was designed by the Shanghai Institute of Applied Physics,Chinese Academy...     

Characterization of molten 2LiF–BeF_2 salt impregnated into graphite matrix of fuel elements for thorium molten salt reactor

The impregnation behavior of molten 2LiF–BeF_2(FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor(TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury intrusion, molten salt impregnation, X-ray diffraction, and scanning electron microscopy techniques.It was found that the entrance pore diameter of the graphite matrix is less than 1.0 μm and the contact angle is about 135°. The threshold impregnation pressure was found to be around 0.6 MPa experimentally, consistent with the predicted value of 0.57 MPa by the Washburn equation. With the increase of pressure from 0.6 to 1.0 MPa, the average weight gain of the matrix increased from 3.05 to 10.48%,corresponding to an impregnation volume increase from 2.74 to 9.40%. The diffraction patterns of FLiBe are found in matrices with high impregnation pressures(0.8 MPa and1.0 MPa). The FLiBe with sizes varying from tens of nanometers to a micrometer mainly occupies the open pores in the graphite matrix. The graphite matrix could inhibit the impregnation of the molten salt in the TMSR-SF with a maximum operation pressure of less than 0.5 MPa.     

Structural investigation of thorium in molten lithium-calcium fluoride mixtures for salt treatment process in molten salt reactor

X-ray absorption fine structure (XAFS) measurements on thorium fluoride in molten lithium-calcium fluoride mixtures and molecular dynamics (MD) simulation of zirconium and yttrium fluoride in molten lithium-calcium fluoride mixtures have been carried out. In the molten state, coordination number of thorium (N_i) and inter ionic distances between thorium and fluorine in the first neighbor (r_i) are nearly constant in all mixtures. However the fluctuation factors (Debye-Waller factor (σ²) and C₃ cumulant) increase until xCaF₂ = 0.17 and decrease by addition of excess CaF₂. It means that the local structure around Th⁴⁺ is disordered until xCaF₂ = 0.17 and stabilized over xCaF₂ = 0.17. The variation of fluctuation factors is related to the number density of F⁻ in ThF₄ mixtures and the stability of local structure around Th⁴⁺ increases with decreasing the number density of F⁻ in ThF₄ mixtures. This tendency is common to those in the ZrF₄ and YF₃ mixtures. However, in the case of YF₃ mixtures, the local structure around Y³⁺ becomes disordered until xCaF₂ = 0.40 and it becomes stabilized by addition of excess CaF₂. The difference between ThF₄ mixtures and YF₃ mixtures is related to the difference of Coulumbic interaction between Th⁴⁺-F⁻ and Y³⁺-F⁻. Therefore, the variation of local structure around cation is related to not only number density of F⁻ in molten salts but also the Coulumbic interaction between cation and anion.     

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(BB) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/~(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PBB) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.     

Radiological environmental impact analysis of a 2-MW thorium molten salt reactor during an accident

The thorium molten salt reactor–liquid fuel(TMSR-LF1) has inherent safety features. The accident occurrence possibility and their consequences are much lower for the TMSR-LF1 than that of traditional reactors.Based on accident analysis, the maximum credible accident and the radioactive source terms of the TMSR-LF1 were first estimated. Then, the total effective dose of the maximum credible accident was calculated. Based on calculations, the cover gas flow rate can significantly affect the radiation consequences of the maximum credible accident when it changes from 0 to 10 L/min. If no cover gas is flowing, a site-area emergency would be required within the range of 50–73 m from the reactor. In the case of cover gas flow, only an abnormal notification and an alert two emergency class would be required within the range of 50 m.     

Reactor protection system testing for the solid fuel thorium molten salt reactor

Safety system testing is one of the most rigorous and time-consuming requirements in the verification and validation process for reactor protection systems(RPSs).This paper presents the development of a test system for the fully digital and field-programmable gate array-based RPS of the solid fuel(SF) thorium-breeding molten salt pebble bed fluoride salt-cooled reactor(TMSR),denoted as the TMSR-SF1 project,developed by the Chinese Academy of Sciences.The test system is applied to the RPS to ensure that it fully meets its designed functions and system specifications.We first introduce the testing principles and methods.Then,the hardware component designs and the software program development of the test system are discussed.Finally,the test process and test results are discussed and summarized.     

氢化锆慢化熔盐堆钍铀转换性能初步分析

中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、~(233)U浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆~(233)U初始浓度降低到2.5×10~(-2)以内;氢化锆慢化熔盐堆在超热谱条件下,其~(233)U初装载量和超铀核素产量较小,同时堆芯较为紧凑。     

Large eddy simulation of unsteady flow in gas-liquid separator applied in thorium molten salt reactor

Axial gas-liquid separators have been adopted in fission gas removal systems for the development of thorium molten salt reactors. In our previous study, we observed an unsteady flow phenomenon in which the flow pattern is directly dependent on the backpressure in a gas-liquid separator; however, the underlying flow mechanism is still unknown. In order to move a step further in clarifying how the flow pattern evolves with a variation in backpressure, a large eddy simulation(LES) was adopted to study the flow field evolution. In the simulation, an artificial boundary was applied at the separator outlet under the assumption that the backpressure increases linearly. The numerical results indicate that the unsteady flow feature is captured by the LES approach, and the flow transition is mainly due to the axial velocity profile redistribution induced by the backpressure variation. With the increase in backpressure,the axial velocity near the downstream orifice transits from negative to positive. This change in the axial velocity sign forces the unstable spiral vortex to become a stable rectilinear vortex.     

Analysis of tritium production in a 2 MW liquid-fueled molten salt experimental reactor and its environmental impact

Tritium release is one of the most concerning topics in nuclear power plants. Here, the tritium production in a 2 MW liquid-fueled molten salt experimental reactor(TMSR-LF1) was calculated by ORIGEN-S with an updated cross-section library generated by TRITON in SCALE 6.1.3 code system. The results show that the tritium production rate and normalized tritium production rate of TMSR-LF1 are 8.90x10~(11)Bq/day and4.45x10~(11)Bq/(MW day), respectively. The environmental impact of tritium was analyzed via PC-CREAM 08 with an assumed 36 % release rate of tritium referring to the molten salt reactor experiment. During normal operations, the maximum tritium concentration is 1.4 Bq/m~3 under normal condition, and the corresponding individual dose to the public is about 1 μSv/a; under extreme conditions, the maximum concentration and corresponding individual doses are 11.8 Bq/m~3 and 9 μSv/a, respectively.Ingestion is the main exposure pathway and accounts for62 % of the total dose. Of this, 35 % is from organically bound tritium.     

一种液态燃料熔盐堆堆芯流量分配设计

在液态燃料熔盐堆(Molten salt reactor,MSR)热工水力设计中,为实现堆芯径向功率展平需对堆芯流量分配进行设计,使得堆芯进口流量分布正比于释热量分布,而下腔室结构和流场分布对堆芯流量分配起决定性作用。利用FLUENT软件对堆芯三维流场进行模拟,通过调节下腔室结构和流量分配装置,对下腔室流场分布进行优化,最终实现堆芯流量合理分配。数值模拟结果表明,喇叭状下腔室比椭球形下腔室熔盐通道流量标准差降低4.2%,设置流量分配板熔盐通道流量标准差降低29.2%;改变下腔室结构和设置流量分配装置能够较好调节流量分配和功率分布匹配性,该结果可为液态熔盐堆堆芯优化设计提供依据。