DYN1D-MSR dynamics code for molten salt reactors

This paper reports about the DYN1D-MSR code development and dynamics studies of the molten salt reactors (MSR) – one of the ‘Generation IV International Forum’ concepts. In this forum the graphite-moderated channel type MSR based on the previous Oak Ridge National Laboratory research is considered.The liquid molten salt serves as a fuel and coolant, simultaneously and causes two physical peculiarities: the fission energy is released predominantly directly into the coolant and the delayed neutrons precursors are drifted by the fuel flow. The drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit and it can lead to a reactivity loss or gain in the case of fuel flow acceleration or deceleration, respectively. Therefore, specific 3D tool based on in house code DYN3D was developed in FZR. The code DYN3D-MSR is based on the solution of two-group neutron diffusion equation by the help of a nodal expansion method and it includes models of delayed neutrons drift and specific MSR heat release distribution.In this paper the development and verification of 1D version DYN1D-MSR of the code is described. The code has been validated with the experimental data gained from the molten salt reactor experiment performed in the Oak Ridge and after the validation it was applied to several typical transients (overcooling of fuel at the core inlet, reactivity insertion, and the fuel pump trip).     

DYN3D-MSR spatial dynamics code for molten salt reactors

The development of spatial dynamics code for molten salt reactors (MSRs) is reported in this paper. The graphite-moderated channel type MSR – one of the ‘Generation IV’ concepts – was selected for the numerical simulation. It has several peculiarities (e.g. the drift of delayed neutrons precursors), which disable the use of standard dynamics codes. Therefore, the own DYN3D-MSR code was developed. It is based on the light water reactor code DYN3D and it allows transients simulation by 3D neutronics and parallel channel thermal-hydraulics. The neutronics and thermal-hydraulics were modified for the MSR peculiarities, where the experience from DYN1D-MSR development was exploited. The code was validated on experimental results from the MSRE experiment done in Oak Ridge National Laboratory and by the comparison with other codes especially with the 1D version. However, by the 3D code transients can be simulated, where space-dependant efforts are relevant, like local blockage of fuel channels or local temperature perturbations.     

Non-proliferation attributes of molten salt reactors

Molten salt reactors (MSR) have many non-proliferation attributes. They can operate on the thorium-uranium fuel cycle which protects the fissile material by the daughter products of the inseparable U-232. MSRs can completely fission all plutonium and HEU, and as desired, ‘convert’ them to U-233. This also results in high, and efficient resource utilization, while diminishing the plutonium stock. On line processing, when applied, could free the waste from all fissile material. The fuel in the reactor stays protected by the intense radiation of the fission products. Fuel can also be protected in the reactor as well as outside the reactor by denaturing with natural uranium. A wide variety of MSRs are available, from ‘once through’ minimum processing reactors to ones with fuel processing which can breed fuel for converters. MSRs are extremely safe and simple reactors with good economic potential.     

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。     

Development of a safety analysis code for molten salt reactors

The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.     

Effect of microconstants uncertainty on transplutonium actinides burnup in molten salt reactors

An evaluation method and results for the error due to microconstants uncertainties in the calculation of neptunium and transplutonium actinide burnup in a molten-salt reactor are presented. The method developed treats the characteristics of a reactor in an equilibrium state and assumes that Np, Am, Cm, and other transplutonium elements as well as material for maintaining criticality are fed continually into the reactor. The perturbation of the equilibrium characteristics of the reactor is described using a linear approximation taking account of the limitations on the prescribed power, k_eff of the reactor, and the actinide content in the fuel salt. The error in the burnup rate is calculated for a homogeneous reactor with NaF-ZrF₄ salt. Different sources are used to determine the errors in the microconstants. The resulting error obtained for Np, Am, Cm, and other transplutonium element burnup ranges from 12 to 51% depending on the reactor power. __________ Translated from Atomnaya énergiya, Vol. 102, No. 5, pp. 270–276, May, 2007.     

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.     

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...     

The influence of Xe-135m on steady-state xenon worth in thermal molten salt reactors

There has been a resurgence of interest in fuel-in-salt Molten Salt Reactors (MSR); a number of governments and private companies are currently undertaking efforts to develop and commercialize MSR technology. Recent nuclear models used in the TENDL nuclear data library have estimated the cross section of the metastable state of ¹³⁵Xe, ^135mXe, to have a much larger cross section than the ground state of ¹³⁵Xe. Thermal MSRs with continual online noble gas stripping of the fuel salt can operate in a regime where ^135mXe makes up a notable fraction of the xenon worth, necessitating the implementation of these new cross-sections in the neutronic analysis of these advanced reactor types. To estimate the effect of ^135mXe on reactor operation, a simplified mathematical model was produced with one neutron energy group and ^135mXe cross section data from the TENDL-2015 nuclear data library. ²³⁵U and ²³³U systems were investigated. It was found that the steady-state xenon reactivity worth was considerably higher for some modes of operation when ^135mXe was included in the xenon worth calculations. Based on available literature, it was found that proposed MSR concepts may operate in the modes of operation where ^135mXe has a notable impact on steady-state xenon worth. This work highlights the need to include ^135mXe in MSR models and the importance of acquiring evaluated cross-sections for ^135mXe.     

规则球床堆熔盐流动压降与对流换热CFD模拟

球床堆复杂的几何结构导致直接建模进行热工水力模拟非常困难,一般使用多孔介质模型简化处理,但多孔介质已有的压降和对流换热公式在熔盐冷却球床中的有效性仍待验证。本文基于固态燃料熔盐堆建立了6 cm直径小球的规则球床模型,给定球床进口熔盐流量和球壳发热功率,模拟了球床内的稳态流动与换热,计算了对应的压降和对流换热系数,并分别得到了球床压降、对流换热Nu随球床内流动Re变化的曲线。对比发现:模拟压降结果与已有公式差异较大,而模拟对流换热Nu结果与已有公式的差异相对较小。结合模拟结果和已有的公式,拟合得到了修正的压降和对流换热Nu公式。将修正公式应用于3 cm直径规则球床中,结果表明多孔介质修正模型与直接模拟结果一致。     

HYSYS软件应用于熔盐冷却系统瞬态分析的可行性研究

系统软件对熔盐堆冷却剂系统的整体模拟、瞬态分析和安全研究起到至关重要的作用。针对化工领域系统仿真软件HYSYS进行二次开发,植入熔盐物性,修改熔盐换热模型,利用软件已有模型(等效回路加热器模型)尝试分析其对熔盐冷却系统仿真的可行性。为验证修改后的软件的可靠性,对中国科学院上海应用物理研究所钍基核能中心硝酸盐熔盐(KNO3-NaNO2-NaNO3 Molten Salt,HTS)实验装置进行了系统仿真模拟,并与实验结果进行了对比。结果显示,扩展后的软件对熔盐冷却系统的分析研究具有较好的适用性。     

熔盐堆堆芯流体力学计算的GPU并行方法研究

使用计算流体力学(Computational Fluid Dynamics,CFD)数值方法对熔盐堆堆芯的流动和热传导等相关物理问题进行模拟求解,需要大量的计算时间。利用图形处理器(Graphics Processing Unit,GPU)加速技术对开源CFD软件Code_Saturne进行二次开发,研究求解熔盐堆堆芯流场的GPU并行算法。采用OpenACC语言在GPU上实现了向量运算、矩阵向量相乘等基本线性代数运算,从而实现预处理共轭梯度法(Preconditioned Conjugate Gradients,PCG)的GPU并行算法,并使用该算法求解压力状态方程。模拟了方腔驱动流模型及带下降段的熔盐堆堆芯模型的流场分布。结果表明,GPU加速后的软件与原版软件的结果一致,但计算时间更少,证明了GPU算法的正确性及有效的加速性。     

新型熔盐快堆中辐照熔盐靶件生产238Pu的物理分析

以氯化物熔盐为靶基质对新型熔盐快堆中²³⁸Pu的生产进行了分析,使用SCALE6.1（Standardized Computer Analyses for Licensing Evaluation Version 6.1）程序,对比了不同靶基质与靶件半径在²³⁸Pu生产中²³⁷Np的转换率与利用率,分析了反射层的能谱分布、不同位置辐照孔道的²³⁷Np反应截面、靶件插入对堆芯反应性的影响以及生成²³⁶Pu杂质浓度,并计算了²³⁸Pu的纯度及产量随辐照时间的变化。结果表明:NpCl₄纯盐靶基质的²³⁷Np转换率较高,减小靶件半径可提高²³⁷Np利用率;远离堆中心位置的辐照孔道热中子份额较高,且靶件插入对堆芯反应影响较小;辐照孔道内靶件的²³⁶Pu浓度可减小至1×10^-7以下,²³⁸Pu纯度超过98%;当辐照周期为40 d时,     

Study on the effect of thermal deformation on the liquid seal of high-temperature molten salt pump in molten salt reactor

The high-temperature molten salt pump is the core equipment in a molten salt reactor that drives the flow of the molten salt coolant. Rotor stability is key to the continuous and reliable operation of the molten salt pump, and the liquid seal at the wear ring can affect the dynamic characteristics of the rotor system. When the molten salt pump is operated in the hightemperature molten salt medium, thermal deformation of the submerged parts inevitably occurs, changing clearance between the stator...     

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.     

Recovery of minor actinides from spent molten salt waste and decontamination of molten salt waste

Recovery of minor actinides from spent molten salt is one of the important issues. Decontamination of spent molten salt waste is also the problem to be solved for establishment of pyrochemical reprocessing. The decontamination method of spent molten salt waste with recovery of minor actinides has been proposed. Our proposed process is based on the hydrometallurgical process. This process consists of the following processes. First, the spent molten salt waste is dissolved in aqueous solution. Next, the minor actinides are recovered by chromatographic techniques using the pyridine resin in the methanolic hydrochloric acid solution. In the last process, the spent molten salt waste is decontaminated by the cation-exchange chromatography. In the present paper, the adsorption behavior of minor actinides, rare earth elements, alkaline earth elements, and alkali metal elements on pyridine resin is reported. The demonstration experiment of the recovery of the minor actinides from simulant spent molten salt waste is also reported.     

Numerical investigation of natural convection characteristics of a heat pipe-cooled passive residual heat removal system for molten salt reactors

The limited availability of studies on the natural convection heat transfer characteristics of fluoride salt has hindered progress in the design of passive residual heat removal systems(PRHRS) for molten salt reactors. This paper presents results from a numerical investigation of natural convection heat transfer characteristics of fluoride salt and heat pipes in the drain tank of a PRHRS. Simulation results are compared with experimental data,demonstrating the accuracy of the calculation methodo...