Neutronics analysis for MSR cell with different fuel salt channel geometries

The neutronic properties of molten salt reactors(MSRs)differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity.Based on the Monte-Carlo N particle transport code,the effects of the size and shape of the fuel salt channel on the neutron physics of an MSR cell are investigated systematically in this study.The results show that the infinite multiplication factor(k?)first increases and then decreases with the change in the graphite cell size under certain fuel volume fraction(FVF)conditions.For the same FVF and average chord length,when the average chord length is relatively small,the k?values for different fuel salt channel shapes agree well.When the average chord length is relatively large,the k?values for different fuel salt channel shapes differ significantly.In addition,some examples of practical applications of this study are presented,including cell selection for the core and thermal expansion displacement analysis of the cell.

Effects of fuel salt composition on fuel salt temperature coefficient(FSTC)for an under-moderated molten salt reactor(MSR)

With respect to a liquid-fueled molten salt reactor(MSR),the temperature coefficient of reactivity mainly includes the moderator temperature coefficient(MTC)and the fuel salt temperature coefficient(FSTC).The FSTC is typically divided into the Doppler coefficient and the density coefficient.In order to compensate for the potentially positive MTC,the FSTC should be sufficiently negative,and this is mostly optimized in terms of the geometry aspect in pioneering studies.However,the properties of fuel salt also directly influence the FSTC.Thus,the effects of different fuel salt compositions including the^(235)U enrichment,heavy metal proportion in salt phase(HM proportion),and the^7Li enrichment on FSTC are investigated from the viewpoint of the essential six-factor formula.The analysis is based on an undermoderated MSR.With respect to the Doppler coefficient,the temperature coefficient of the fast fission factors(a_T(ξ))is positive and those of the resonance escape probability(a_T(p)),thermal reproduction factor(a_T(g)),thermal utilization factor(a_T(f)),and total non-leakage probability(a_T(λ))are negative.With respect to the density coefficient,a_T(p)and a_T(g)are positive,while the others are negative.The results indicate that the effects of the^(235)U enrichment and HM on FSTC are mainly reflected in a_T(e)and a_T(p),which are the dominant factors when the neutron spectrum is relatively hard.Furthermore,the^7Li enrichment influences FSTC by a_T(f)and a_T(λ),which are the key factors in a relative soft spectrum.In order to obtain a more negative FSTC for an under-moderated MSR,the possible positive density coefficient,especially its a_T(p),should be suppressed.Thus,a lower^(235)U enrichment(albeit higher than a certain value,5 wt%in this article)along with a lower HM proportion and/or a higher^7Li enrichment are recommended.The analyses provide an approach to achieve a more suitable fuel salt composition with a sufficiently negative FSTC.

Plutonium utilization in a small modular molten-salt reactor based on a batch fuel reprocessing scheme

A molten salt reactor(MSR)has outstanding features considering the application of thorium fuel,inherent safety,sustainability,and resistance to proliferation.However,fissile material^(233)U is significantly rare at the current stage,thus it is difficult for MSR to achieve a pure thorium-uranium fuel cycle.Therefore,using plutonium or enriched uranium as the initial fuel for MSR is more practical.In this study,we aim to verify the feasibility of a small modular MSR that utilizes plutonium as the starting fuel(SM-MSR-Pu),and highlight its advantages and disadvantages.First,the structural design and fuel management scheme of the SM-MSR-Pu were presented.Second,the neutronic characteristics,such as the graphite-irradiation lifetime,burn-up performance,and coefficient of temperature reactivity were calculated to analyze the physical characteristics of the SM-MSR-Pu.The results indicate that plutonium is a feasible and advantageous starting fuel for a SM-MSR;however,there are certain shortcomings that need to be solved.In a 250 MWth SM-MSR-Pu,approximately 288.64 kg^(233)U of plutonium with a purity of greater than 90% is produced while 978.00 kg is burned every ten years.The temperature reactivity coefficient decreases from -4.0 to -6.5 pcm K^(-1) over the 50-year operating time,which ensures a long-term safe operation.However,the amount of plutonium and accumulation of minor actinides(MAs)would increase as the burn-up time increases,and the annual production and purity of^(233)U will decrease.To achieve an optimal burn-up performance,setting the entire operation time to 30 years is advisable.Regardless,more than 3600 kg of plutonium eventually accumulate in the core.Further research is required to effectively utilize this accumulated plutonium.

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(B&B) 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(PB&B) 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.

Analysis of burnup performance and temperature coefficient for a small modular molten‑salt reactor started with plutonium

In a thorium-based molten salt reactor(TMSR),it is difficult to achieve the pure 232Th–^(233)U fuel cycle without sufficient^(233)U fuel supply.Therefore,the original molten salt reactor was designed to use enriched uranium or plutonium as the starting fuel.By exploiting plutonium as the starting fuel and thorium as the fertile fuel,the high-purity^(233)U produced can be separated from the spent fuel by fluorination volatilization.Therefore,the molten salt reactor started with plutonium can be designed as a^(233)U breeder with the burning plutonium extracted from a pressurized water reactor(PWR).Combining these advantages,the study of the physical properties of plutonium-activated salt reactors is attractive.This study mainly focused on the burnup performance and temperature reactivity coefficient of a small modular molten-salt reactor started with plutonium(SM-MSR-Pu).The neutron spectra,^(233)U production,plutonium incineration,minor actinide(MA)residues,and temperature reactivity coefficients for different fuel salt volume fractions(VF)and hexagon pitch(P)sizes were calculated to analyze the burnup behavior in the SM-SMR-Pu.Based on the comparative analysis results of the burn-up calculation,a lower VF and larger P size are more beneficial for improving the burnup performance.However,from a passive safety perspective,a higher fuel volume fraction and smaller hexagon pitch size are necessary to achieve a deep negative feedback coefficient.Therefore,an excellent burnup performance and a deep negative temperature feedback coefficient are incompatible,and the optimal design range is relatively narrow in the optimized design of an SM-MSR-Pu.In a comprehensive consideration,P=20 cm and VF=20%are considered to be relatively balanced design parameters.Based on the fuel off-line batching scheme,a 250 MWth SM-MSR-Pu can produce approximately 29.83 kg of ^(233)U,incinerate 98.29 kg of plutonium,and accumulate 14.70 kg of MAs per year,and the temperature reactivity coefficient can always be lower than−4.0pcm/K.

FLiNaK熔盐中CsF的定向凝固分离研究

熔盐反应堆采用氟化物熔盐作为冷却剂和燃料载体,运行后的燃料成分为铀、钍、裂变产物和载体盐,对裂变产物进行分离并回收有效组分复用,可以提高反应堆的运行经济性,并且使放射性废物最小化。定向凝固技术是利用多元混合物的相平衡特性和凝固过程元素迁移机制实现物质分离,具有工艺操作简单、无副产物产生等优点,有望用于燃料盐中裂变产物分离。在自制的冷棒式定向凝固实验装置上,研究了FLiNaK熔盐体系内典型裂变产物CsF在不同工艺条件下定向凝固后的含量分布。研究结果表明:通过控制冷却凝固速度,得到的凝固盐中Cs元素的含量在径向上呈现出梯度分布,由内向外依次递减,外侧凝固盐中Cs含量相较液相中最高降低约90%,表明燃料盐中裂变产物定向凝固分离具有一定的可行性。

熔盐堆核燃料盐贮存的核临界安全分析

熔盐堆是国际公认推荐的6种第四代反应堆型之一,可以使用液态核燃料,其核燃料生产、转运和贮存所涉及工艺过程与常规固态核燃料堆型也不同。为做好核燃料管理和核安全监管,有必要对其贮存的临界安全进行分析。本研究参考美国液态燃料熔盐反应堆MSRE(Molten Salt Reactor Experiment)相关设计参数,通过选取液态燃料熔盐堆核燃料的贮存建模、临界参数分析、蒙特卡罗中子输运软件仿真模拟计算,分析不同因素对核燃料盐贮存的影响,总结了设计模型下干燥环境贮存、水淹环境贮存的keff值及与燃料盐总质量变化的规律。最终,得到了不同情况下次临界安全控制的质量及与对应原料盐、中间产物、考虑容器壁影响时的对比。本研究结合法律法规及核材料流转过程进行分析讨论,归纳核燃料盐核临界安全特性,从核安全监管角度首次提出了相关监督审评要点。

双面冷却熔盐堆组件热工分析程序开发与验证

新型液态燃料熔盐堆双面冷却组件相比传统熔盐堆组件具有更大的换热面积和更短的石墨导热距离,因而具有更低的石墨温度热点,然而由于其不同的几何结构特征和熔盐自发热特性,组件内存在独特的热量分配和冷却剂流量分配的问题。为准确高效评估新型组件的热安全边界,有必要开发新的计算分析工具。本研究基于MATLAB开发了适用于熔盐堆双面冷却组件的一维稳态热工分析程序THDA-MSR(Thermal-Hydraulic Analysis Code for Dual Cooled Assembly-Molten Salt Reactor),考虑熔盐自发热的特点,建立了组件内外流道内熔盐的一维温度分布模型和熔盐石墨对流换热模型以及石墨导热模型,根据并联通道压损相等原则建立流量分配模型,通过CFD(Computational Fluid Dynamics)数值模拟对程序计算结果进行了验证,并初步分析了各结构参数对组件最大熔盐出口温度和石墨温度热点的影响。THDA-MSR的计算结果与CFD结果吻合较好,压损偏差小于4.84%,石墨温度热点偏差小于0.15%,分析结果发现:外通道截面积占总流道截面积比是影响最大熔盐出口温度和石墨温度热点的关键参数。以上研究结果表明,自编程序能很好地预测组件流量分布、组件压损、熔盐温度分布和石墨温度热点。THDA-MSR能够应用于熔盐堆双面冷却组件热工水力性能分析评估,对于组件选型设计有较大工程参考价值。

工作条件对干法后处理高温熔盐离心泵影响的数值模拟研究

本文针对乏燃料干法后处理高温熔盐离心泵输送LiCl-KCl熔盐过程,开展了不同操作条件和物性参数下的数值模拟研究。基于Fluent数值模拟软件,选用标准k-ε湍流模型,压力和速度耦合采用Coupled算法,对比了熔盐泵水力性能和内部流场的变化情况。结果表明,不同工作条件下熔盐泵外特性曲线走向及内部流场压力、速度分布规律相一致。等流量工况下转速增大熔盐泵扬程、功率增大,最佳效率点向大流量偏移,推荐干法熔盐泵转速范围为2750~3150 rpm。泵的扬程、效率均随输送熔盐粘度的增大而减小。泵的扬程对输送LiCl-KCl熔盐密度的变化不敏感,随密度增大,泵的功率和效率均上升,相同条件下熔盐粘度对泵性能的影响大于密度。本文结果可为干法后处理高温熔盐泵设计和实验的改进提供参考。

干法后处理熔盐电解精炼过程数学模型研究

熔盐电解精炼是乏燃料干法后处理的核心工艺单元,通过数学模型探索高温熔盐电解精炼过程的化学与电化学变化,可为电解精炼工艺优化和设备设计提供参考依据。本文基于电化学热力学及物质传递公式建立了乏燃料熔盐电解精炼过程的数学模型,以铀钚锆三元合金燃料为研究对象,计算了燃料中关键元素的电极电势、分电流及物料分布随时间的变化。采用向后差分法对物料分布变化方程进行离散,通过文献实验数据对建立的数学模型进行了准确性验证。结果表明,模拟计算所得阴极沉积铀产品与实验数据的相对误差为2.80%,所建数学模型具有较好的拟合性。同时采用所建模型模拟计算了电流强度对乏燃料电解精炼过程的影响,结果表明电解速率与电流强度呈正比,不改变钚铀锆的溶解和沉积顺序。

氧化物乏燃料锂热还原技术研究进展

以电解精炼为主流的干法后处理技术虽处于实验室研究阶段,但具有可处理最广泛的燃料种类以及反应体系稳定性更高等独特优势。在电解精炼步骤前需通过电还原或热还原将氧化物乏燃料还原为金属。热还原需要使用钙、锂和镁等作为还原剂。相比其他碱金属还原,锂热还原具有回收方法简单、所需温度适中以及设备要求较低等优点,在国内外已相继进行了各方面工艺条件以及理论的研究。调研了氧化物乏燃料锂热还原研究相关进展,对比了各国锂热还原工艺的特点,结合热力学数据分析了高温锂还原的相关机理,从动力学角度对反应的影响因素进行了讨论。总结了现有研究的不足,并对未来锂热还原的趋势进行了展望,为我国干法后处理中工艺流程的研究提供参考。

二元碳酸熔盐在氧化镍表面热物性的分子动力学模拟

采用分子动力学模拟方法,研究了高温下碳酸钠、碳酸钾二元熔盐与氧化镍板的界面体系。研究得到了界面热阻、热导率和黏度等热物性随温度的变化规律,对比了不同温度下均相熔盐材料和界面处熔盐材料的热物性的差异,并且通过密度分布和径向分布函数揭示了热物性发生变化的微观机制。模拟结果显示:温度升高时,熔盐离子间距离增加,范德华力和库仑力相互作用减弱,使得离子间能量传递更加困难,界面热阻、熔盐热导率均下降;同时,熔盐黏度也因为相互作用减弱,熔盐离子自身振动能量增加、运动趋向增强而下降。

秦山核电站海域有害盐在带温核级材料表面沉积实验设计

含Cl有害盐在服役构件表面的沉积量,是影响服役构件腐蚀进程的重要因素。秦山核电站临海而建,面临含Cl有害盐沉积引起的腐蚀问题。该文通过实地环境调研,并根据调研结果开展实验室盐雾沉积实验设计,结果表明:以ASTM D1141-98(2021版)标准中的人工海水Cl元素含量为基准,将秦山核电站海域海水各化合物含量乘以4.075 26进行放大,可得到符合ASTM标准设计的盐雾沉积用有害盐模拟溶液成分。之后,该文进一步开展90℃带温核级材料表面有害盐沉积实验,探究临海服役环境下,有害盐在秦山核电站乏燃料贮罐材料(带温核级材料)表面的沉积规律,为开展实际服役环境下的核电站材料服役寿命评估、服役性能评价提供一种实验室设计思路与借鉴。

面向乏燃料干法后处理的减压蒸馏回收高纯金属研究进展

在乏燃料干法后处理中,通常使用减压蒸馏纯化阴极产物以及回收熔盐。本文围绕减压蒸馏在乏燃料干法后处理中的应用,系统地总结了该技术的总体进展,指出熔盐堆燃料处理过程中减压蒸馏技术将来需要重点解决的问题是高温、高放射性环境下的设备耐受性问题,与其他分离技术的结合,形成更加完善的熔盐堆燃料处理体系是将来的发展趋势。综述了当前熔盐系统的选择与优化、蒸馏设备与参数的优化方面的研究进展,指出减压蒸馏技术将来发展的重点是进一步优化熔盐系统,探究不同熔盐体系对于分离乏燃料中各种元素的适应性,并优化蒸馏设备的设计及参数,以提高分离效率和回收率。

航煤加氢装置原料/反应产物换热器铵盐结晶原因分析及应对措施

某石化公司190×10^(4)t/a航煤加氢装置开工两年后原料/反应产物换热器管束铵盐结晶造成换热效率下降,换热器管程出口温度下降,出入口压差逐步上升,针对原料/反应产物换热器管束铵盐结晶问题,通过对比原料性质、工艺流程,运行条件,对铵盐结晶形成的原因及过程进行深入分析,提出增加原料/反应产物换热器注水线间断注水,取得了控制铵盐结晶较好的效果。

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 Th-U breeding capability for an accelerator-driven subcritical molten salt reactor

Accelerator-driven systems based on molten salt fuel have several unique advantages and features for advanced nuclear fuel utilization.The aim of this work was to study the Th-U breeding capability in such systems,known as‘‘accelerator-driven subcritical molten salt reactors’’(ADS–MSRs).Breeding capacities including conversion ratio and net^(233)U production for various subcriticalities and different minor actinides(MA)loadings were analyzed for an ADS–MSR.The results show that the subcriticality of the core has a considerable effect on the Th-U breeding.A high subcriticality is favorable to improving the conversion ratio,increasing the net^(233)U production,and reducing the doubling time.Specifically,the doubling time for k_(eff)of 0.99 is larger than 80 years,while the counterpart for k_(eff)of 0.93 is only approximately22 years.Nevertheless,in an ADS–MSR with a high initial MA loading,MA results in a non-negligible^(233)U depletion in the first two decades,while increasing the net^(233)U production compared to reactors without MA loading.During the 50 years of operation,for the subcritical reactor(k_(eff)0:97)with MA fraction increasing from 1 to 14%,the net^(233)U production increases from 3.94 to 8.24 t.

Advance on Molten Carbonate Fuel Cell and Research on Some Key Problems

The paper is a summary of researches on molten carbonate fuel cell. On the same time, several key technology difficulties are discussed. Combining with our recent studies, the accessements to these problems are given out and they will be references for future works.

Effect of 37Cl enrichment on neutrons in a molten chloride salt fast reactor

A molten chloride salt fast reactor(MCFR)is well suited to fuel breeding and the transmutation of transuranium(TRU)elements owing to its advantageous features of fast neutron spectrum and high TRU solubility.However,the neutron absorption cross section of 35Cl is approximately 1000 times greater than for 37Cl,which has a significant impact on the neutron physical characteristics of a MCFR.Based on an automatic online refueling and reprocessing procedure,the influences of 37Cl enrichment on neutron economy,breeding performance,and the production of harmful nuclides were analyzed.Results show that 37Cl enrichment strongly influences the neutron properties of a MCFR.With natural chlorine,233U breeding cannot be achieved and the yields of S and 36Cl are very high.Increasing the 37Cl enrichment to 97%brings a clear improvement in its neutronics property,making it almost equal to that corresponding to 100%enrichment.Moreover,when 37Cl is enriched to 99%,its neutronics parameters are almost the same as for 100%enrichment.Considering the enrichment cost and the neutron properties,a 37Cl enrichment of 97%is recommended.Achieving an optimal neutronics performance requires 99%37Cl enrichment.

熔盐堆堆芯应急排盐系统可靠性分析及优化

堆芯应急排盐系统作为熔盐堆特有的安全系统,具有排盐和余热排出功能,为熔盐堆提供了一种紧急停堆方式。为定量化分析堆芯应急排盐系统的可靠性,以美国橡树岭实验室的熔盐实验堆(Molten Salt Reactor Experiment,MSRE)为研究对象,使用概率安全分析软件RiskSpectrum建立和计算MSRE堆芯应急排盐系统故障树,得到系统失效概率为5.62×10^(-4),并进行最小割集分析和重要度分析,识别出影响系统失效的关键因素是外套管泄漏失效、控制冷冻阀冷却气的电磁阀共因失效和气动阀共因失效。通过套管换热元件中减少使用焊缝连接,以及采用不同类型部件控制冷冻阀冷却气,可明显降低系统失效概率。分析结果为液态熔盐堆应急排盐系统工程应用研究提供参考。