Improvements and validation of the transient analysis code MOREL for molten salt reactors

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective.     

钍基液态燃料熔盐实验堆氙毒分析

熔盐堆作为第四代核能系统堆型之一,液态燃料形态的特点使其可以实现在线处理和在线添料。为了提高中子经济性可以利用在线处理的氦鼓泡法,将氦气通入反应堆一回路,去除堆芯内的裂变气体(如Xe、Kr)。基于钍基熔盐液态堆(Thorium Molten Salt Reactor-Liquid Fuel1,TMSR-LF1)概念设计,结合熔盐实验堆(Molten Salt Reactor Experiment,MSRE)氙毒模型,分析了鼓泡法去除氙毒中~(135)Xe扩散规律和去除效率对氙毒的影响,并给出了对应的初始有效增殖因子的变化规律。分析结果表明,虽然存在~(135)Xe会大量向石墨扩散的可能性,但是鼓泡法仍然可以有效去除TMSR-LF1堆芯内的~(135)Xe,减小堆芯毒性,提高反应性。     

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...     

Steady thermal hydraulic analysis for a molten salt reactor

The Molten Salt Reactor (MSR) can meet the demand of transmutation and breeding. In this study, theoretical calculation of steady thermal hydraulic characteristics of a graphite-moderated channel type MSR is conducted. The DRAGON code is adopted to calculate the axial and radial power factor firstly. The flow and heat transfer model in the fuel salt and graphite are developed on basis of the fundamental mass, momentum and energy equations. The results show the detailed flow distribution in the core, and the temperature profiles of the fuel salt, inner and outer wall in the nine typical elements along the axial flow direction are also obtained.     

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%。     

Analysis of heat source distribution in internal circulating surface heat transfer molten salt reactor

The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods.     

熔盐球床堆堆芯入口热工水力特性数值分析

堆芯入口流场设计是小型固态燃料熔盐堆系统项目内容之一,它对反应堆结构的稳定性、堆芯温度和流场分布有着非常重要的影响。研究了熔盐流道流通面积变化对堆芯入口温度、流场分布及压降的影响,优化熔盐流道几何结构。以小型熔盐球床堆模型为研究对象,取符合实际边界条件的输入参数,通过改变熔盐流道流通面积,使用计算流体力学(Computational Fluid Dynamics,CFD)通用程序Fluent 16.0对堆芯入口内熔盐的热工水力特性进行数值模拟。在考虑实际下反射层流道的流通面积占比最大为18.14%下,研究了熔盐流道流通面积占比在区间[0,15.00%]变化。结果表明,堆芯活性区熔盐最高局部热点温度随熔盐流道流通面积比的增大而增高;堆芯入口内的压降随下反射层熔盐流道流通面积比的减小而增大;在径向方向上流进孔道的熔盐流速随着孔道远离堆芯位置而增大。本研究可为小型固态燃料球床熔盐堆优化设计提供一定的参考价值。     

Properties of phosphate glass waste forms containing fluorides from a molten salt reactor

Radioactive fluoride wastes are generated during the operation of molten salt reactors(MSRs) and reprocessing of their spent fuel.Immobilization of these wastes in borosilicate glass is not feasible because of the very low solubility of fluorides in this host.Alternative candidates are thus an active topic of research including phosphatebased glasses,crystalline ceramics,and hybrid glass-ceramic systems.In this study,mixed fluorides were employed as simulated MSRs waste and incorporated into sodium aluminophosphate glass to obtain phosphate-based waste form.These waste forms were characterized by X-ray diffraction,Raman spectroscopy,and scanning electron microscopy.Leaching tests were performed in deionized water using the product consistency test A method.This study demonstrates that up to 20 mol%of simulated radioactive waste can be introduced into the NaA1 P glass matrix,and the chemical durability is much better than that of borosilicate.The addition of Fe_2O_3 in the NaAlP glass matrix results in increases of the chemical durability at the expense of fluoride loading(to 6.4 mol%).Phosphate glass vitrification of radioactive waste containing fluorides is a potential method to treat and dispose of MSR wastes.     

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.     

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。     

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。     

The investigation of thermal neutron scattering data for molten salt Flibe

Based on the dynamics of Flibe (2LiF-BeF₂), the thermal neutron scattering library (TSL) is generated by NJOY in this paper. We worked on the TSL data through the microscopic analysis. With the increas of temperature and translational weight of liquid Flibe, the inelastic scattering cross section decreases and the average energy of secondary neutron increases. Using TSL of Flibe, the critical calculation for molten salt reactor is also conducted by MCNP. Result shows the k _eff is influenced by thermal neutron scattering obviously (about 0.8%). It is suggested that thermal neutron data should be taken into account in molten salt reactor.     

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.     

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.     

基于球床密实实验装置的熔盐堆分流板设计

在熔盐球床堆设计中,为实现堆芯内部燃料球堆积结构的稳定性,需保证堆芯内部的流场均匀分布。研究基于模拟熔盐球床堆堆芯水力特性的球床密实实验装置(Pebble bed dense experiment facility,PBDE),通过设计不同形状和不同孔道分布的分流板,运用计算流体力学(Computational fluid dynamics,CFD)方法使用FLUENT软件对其堆芯内部的流场分布进行数值模拟,目的是保证实验中堆芯的流场分布均匀稳定。模拟结果表明,平板形分流板较锥形分流板能更好地使堆芯内流场均匀分布;且增加分流板的孔道数目或减小孔径能使堆芯内部的流场更加均匀稳定;比较设计的6种分流板模拟结果,最终给出满足PBDE堆芯流场均匀分布的分流板,为PBDE实验提供了基础,也为熔盐球床堆的堆芯流量分配设计提供技术方案与选型参考。     

熔盐冷冻壁形成及控制实验研究

乏燃料干法后处理中高温熔盐化学工艺过程中存在设备腐蚀问题,以凝固盐层作为容器保护内衬的熔盐冷冻壁技术被认为是一种可行的解决方式。为开展该技术研究,自行研制并搭建了一套硝酸盐(40.0-7.0-53.0wt%Na NO2-Na NO3-KNO3)冷冻壁技术研究实验装置。该装置上熔盐运行温度150-250oC,熔盐最大流量500 L·h-1,循环导热油运行温度5-120oC,导热油流量1.5-15 m3·h-1。目前在该装置上开展了冷冻壁静态形成及平衡维持等工艺研究,实验中采用容器外壁循环导热油冷却换热实现冷冻壁的形成及维持,并试验了冷冻壁技术在熔盐静态工况下应用的工艺条件。冷冻壁形成平均速率可控制在0.2-0.5 mm·min-1。在冷冻壁静态形成过程中,随厚度增大,热交换量逐渐减小,同时冷冻壁层温差逐渐增大,并均呈衰减趋势变化;处于平衡维持状态时,径向温度分布、热流量及冷冻壁厚度均保持稳定,熔盐发热功率即为平衡状态时的热流量,其大小同时与外壁导热油的冷却热流量相等;实验还获得了较理想的静态应用工艺操作条件,为氟化物熔盐冷冻壁的研究及应用积累了经验。     

Numerical study on rotordynamic coefficients of the seal of molten salt pump

Molten salt pump is applied to pump high-temperature molten salt as the primary coolant of a molten salt reactor.The pump,generally a vertical rotor system,suffers from radial force generated by the liquid seal component,and the rotordynamic characteristics of the pump are affected considerably.In this paper,the rotordynamic coefficients of the tooth-on-stator liquid seal in molten salt pump are studied.The flow in the seal region is simulated using computational fluid dynamics technique.Parameters of the inlet loss and the pre-swirl at the inlet region of the seal are calculated.The coefficients of resistance and the wall parameters are obtained from the simulation results by data fitting method.The rotordynamic coefficients are analyzed based on the bulk-flow model of liquid seal.The rotordynamic characteristics,with and without the liquid seal,of the objective molten salt pump are inspected.The first critical speed of the rotor is found to increase.Harmonic analysis shows that the pump,being sensitive to unbalance force though,can operate safely under its design specifications.     

The advanced high-temperature reactor: High-temperature fuel, liquid salt coolant, liquid-metal-reactor plant

The Advanced High-Temperature Reactor is a new reactor concept that combines four existing technologies in a new way: (1) coated-particle graphite-matrix nuclear fuels (traditionally used for helium-cooled reactors), (2) Brayton power cycles, (3) passive safety systems and plant designs from liquid-metal-cooled fast reactors, and (4) low-pressure liquid-salt coolants with boiling points far above the maximum coolant temperature. The new combination of technologies enables the design of a large [2400- to 4000-MW(t)] high-temperature reactor, with reactor-coolant exit temperatures between 700 and 1000°C (depending upon goals) and passive safety systems for economic production of electricity or hydrogen. The AHTR [2400-MW(t)] capital costs have been estimated to be 49 to 61% per kilowatt (electric) relative to modular gas-cooled [600-MW(t)] and modular liquid-metal-cooled reactors [1000-MW(t)], assuming a single AHTR and multiple modular units with the same total electrical output. Because of the similar fuel, core design, and power cycles, about 70% of the required research is shared with that for high-temperature gas-cooled reactors.     

Conceptual design and analysis of a passive residual heat removal system for a 10 MW molten salt reactor experiment

A conceptual design of a passive residual heat removal system was developed for a 10 MW molten salt reactor experiment (MSRE) designed by Oak Ridge National Laboratory (ORNL). The principle, main components and design parameters of the system were presented, and thermal-hydraulic behaviors, such as natural circulation and heat removal ability, were numerically analyzed in the code of C++, especially for the bayonet cooling thimbles. The results show that the system can effectively remove decay heat in the molten salt in an MSRE and has a heat removal rate that approximates to the decay heat generation rate, thus causing the temperature of the molten salt to decrease steadily. The width of the gas gap in the bayonet cooling thimbles has little effect on either the heat exchange or the natural circulation inside the thimbles, while the width of the steam riser, in spite of its slight effect on the heat transfer of the system, greatly influences the natural circulation. With the width of the steam riser increase from 3.6 to 5.1 mm, the mass flow rate increases from 1.9 kg/s to 4.79 kg/s. Finally, three operational schemes were proposed for the passive residual heat removal system, among which that of reducing the bayonet cooling thimbles by three-quarters had the best comprehensive performance.