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.     

基于MOREL2.0对石墨慢化熔盐堆的稳态分析

熔盐堆采用熔融的氟化盐混合物作为燃料和堆芯的冷却剂,由于燃料的流动,熔盐堆在中子学和热工水力学方面与传统固体燃料反应堆有着较大区别。本文基于熔盐堆分析程序MOREL2.0对钍基熔盐堆（TMSR）初步堆芯设计方案进行了稳态计算分析,结果表明：燃料流动对缓发中子先驱核的分布影响较大,并导致169 pcm反应性损失;随燃料在外部回路中滞留时间的增加,k_eff降低,80 s后趋于平稳;TMSR具有负的入口燃料温度系数,具有固有安全性。     

液态熔盐堆运行安全特性初步研究

液态燃料反应堆与固态燃料反应堆相比,原理上有较大不同。液态熔盐堆中由于燃料流动带走缓发中子先驱核在堆外衰变导致堆芯反应性降低,且裂变产物在堆外回路中衰变也会引起一回路发热。本文使用熔盐堆中子动力学程序Cinsf1D探讨2 MW熔盐堆的临界动力学特性和安全特性,研究零功率临界下不同熔盐流速启泵和停泵导致的缓发中子先驱核流失所需改变的控制棒棒位。同时还计算了2 MW恒定功率情况下稳态运行及降低流速时一回路温度分布,并模拟了2 MW额定功率下停泵事件。停泵后由于缓发中子损失减少反应堆功率先缓慢增加,然后迅速降低到接近余热水平。停泵后堆芯温度缓慢增加后稳定在安全值以内,说明熔盐堆具有本征安全性。     

石墨慢化通道式熔盐堆有效缓发中子份额计算方法研究

有效缓发中子份额(βeff)是研究反应堆动力学特性的关键参数。在液态燃料熔盐堆(MSR)中,燃料流动引起缓发中子先驱核(DNP)在堆内的再分布,并使部分DNP在堆外回路衰变,从而导致βeff的计算方法与固态燃料反应堆不同。为评估石墨慢化通道式熔盐堆内燃料流动引起的反应性损失,研究缓发中子随燃料的流动行为,同时为堆设计和安全分析提供依据,分别基于解析方法和数值方法推导了计算βeff的数学模型,计算了熔盐实验堆(MSRE)在额定工况下的DNP损失份额和堆内DNP浓度分布,并分析了燃料在堆外流动时间和入口流量对βeff的影响。结果表明：两种方法均可对DNP行为提供合理描述;固定燃料在堆外流动时间,βeff随入口流量的增加而减小;固定入口流量,βeff随燃料在堆外流动时间的增加而减小,80 s后趋于稳定。     

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.     

物理-热工耦合计算方法在熔盐堆稳态分析中的应用

针对美国橡树岭国家实验室（ORNL）熔盐堆（MSR）实验的堆芯设计,采用物理分析程序MCNP进行三维堆芯功率分布计算。针对以石墨作为慢化剂的堆芯结构,开发了并联多通道程序来进行堆芯热工水力分析。在此基础上,把物理和热工分析程序进行耦合,用ORNL技术报告中的相关内容来验证物理 热工耦合分析的可行性和准确性。结果表明,本工作的耦合计算方法可获得熔盐堆堆芯功率分布、温度分布、压降和流量分配。熔盐堆耦合程序的研发对熔盐堆概念设计、运行分析有重要意义。     

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。     

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。     

Impact of photoneutrons on reactivity measurements for TMSR-SF1

The solid-fueled thorium molten salt reactor(TMSR-SF1) is a 10 MW_(th) test reactor design to be deployed in 5-10 years by the TMSR group.Its design combines coated particle fuel and molten FLiBe coolant for great intrinsic safety features and economic advantages.Due to a large amount of beryllium in the coolant salt,photoneutrons are produced by(y,n) reaction,hence the increasing fraction of effective delayed neutrons in the core by the photoneutrons originating from the long-lived fission products.Some of the delayed photoneutron groups are of long lifetime,so a direct effect is resulted in the transient process and reactivity measurement.To study the impact of photoneutrons for TMSR-SF1,the effective photoneutron fraction is estimated using k-ratio method and performed by the Monte Carlo code(MCNP5) with ENDF/B-Ⅶ cross sections.Based on the coupled neutronphoton point kinetics equations,influence of the photoneutrons is analyzed.The results show that the impact of photoneutrons is not negligible in reactivity measurement.Without considering photoneutrons in on-line reactivity measurement based on inverse point kinetics can result in overestimation of the positive reactivity and underestimation of the negative reactivity.The photoneutrons also lead to more waiting time for the doubling time measurement.Since the photoneutron precursors take extremely long time to achieve equilibrium,a "steady" power operation may not directly imply a "real" criticality.     

石墨慢化通道式熔盐堆的稳态热工水力计算模型

针对石墨慢化通道式熔盐堆的堆芯结构,基于COMSOL Multiphysics程序和MATLAB程序建立了堆芯稳态热工水力学计算模型。该模型对堆芯内固体区域的温度分布采用三维热传导方程进行模拟,对通道内熔盐温度采用一维单相流体模型进行计算。固体区域与熔盐通过熔盐通道壁面的对流换热边界建立热耦合。该模型基于平行通道压力损失相等的原则,分配堆芯内各熔盐通道的流量。通过对比RELAP5程序的计算结果,验证了模型对温度和流量分配计算的正确性。针对2 MWt 液态燃料熔盐堆的一种概念设计,分析了堆芯内三维温度分布和通道间流量分配。该模型可精确计算通道式熔盐堆堆芯内稳态温度分布和流量分配,对堆芯的热工水力学设计具有重要意义。     

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。     

钍基熔盐堆核能系统中熔盐的蒸馏纯化与分离

氟盐具有化学与辐射稳定性高、热容量大、传热性能好、运行温度高和蒸汽压低等优点,被用作熔盐堆的燃料载体和冷却剂。随着熔盐堆技术的发展,开发熔盐的净化、回收工艺非常必要。熔盐减压蒸馏技术基于物质挥发性差异进行组分分离,由于过程操作简单、不引入新的物质等特点,在燃料处理过程中有广泛应用。利用减压蒸馏技术对钍基熔盐堆核能系统的载体盐回收、电解产物纯化、模拟燃料球去除浸渗熔盐等方面进行了研究。研究结果表明,含CsF、SrF_2、LaF_3和ThF_4的FLiNaK盐经减压蒸馏处理,可从FLiNaK中除去SrF_2和LaF_3,去污因子分别为4.4×10~3和1.9×10~3,Th的去污因子为94;通过蒸馏可去除电解产物表面夹带的氟盐,纯化电解产物;燃料球中浸渗熔盐在1 085℃下处理37h可去除石墨球中94%的浸渗熔盐。     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

Calculation of the VVER-1000 coolant transient benchmark using the coupled code systems DYN3D/RELAP5 and DYN3D/ATHLET

Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.     

Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations

Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and three out of four MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical MCP characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. For an adequate modelling of the redistribution of the coolant flow in the reactor pressure vessel during the transient a simplified mixing model for the DYN3D/ATHLET code was developed and validated against a computational fluid dynamics calculation.

The results of both coupled code calculations are in good agreement with the available experimental data. The discrepancies between experimental data and the results of both coupled code calculations do not exceed the accuracy of the measurement data. This concerns the initial steady-state data as well as the time histories during the transient. In addition to the validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermal hydraulic models of the system codes RELAP5 and ATHLET and to explain differences between the calculation results.     

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.     

Analysis on reactivity initiated transient from control rod failure events of a molten salt reactor

In a molten salt reactor （MSR）, the fuel is dissolved in fluoride salt. In this paper, the reactivity worth and reactivity initiated transient of Molten-Salt Reactor Experiment （MSRE） in the control rod failure events are analyzed, The point kinetic coupling heat-transfer model with decay character of six-group delayed neutron precursors due to the fuel motion is applied. The relative power and temperature transient under reactivity step and ramp initiated at different power levels are studied. The results show that the reactor power and temperature increase to a maximum, where they begin to decrease to stable values. Comparing with full power level, the transient result at low power level is more serious. The results are of help in our study on safety characteristics of an MSR system.     

Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels

The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF₂, LiF, ZrF₄ and Li₂BeF₄ eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.     

Development of a steady state analysis code for a molten salt reactor

The molten salt reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for transmutation, and production of electricity, hydrogen and fissile fuels. In this study, a single-liquid-fueled MSR is designed for conceptual research, in which no solid material is present in the core as moderator, except for the external reflector. The fuel salt flow makes the MSR neutronics different from that of conventional reactors using solid fuels, and couples the flow and heat transfer strongly. Therefore, it is necessary to study the core characteristics with due attention to the coupling among flow, heat transfer and neutronics. The standard turbulent model is adopted to establish the flow and heat transfer model, while the diffusion theory is used for the neutronics model, which consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six groups of delayed neutron precursors. These two models which are coupled through the temperature and heat source are coded in a microcomputer program. The distributions of the velocity, temperature, neutron fluxes, and delayed neutron precursors under the rated condition are obtained. In addition, the effects of the inflow temperature, inflow velocity, and the fuel salt residence time out of the core are discussed in detail. The results provide some valuable information for the research and design of the new generation molten salt reactors.