Dynamic simulation of a space gas-cooled reactor power system with a closed Brayton cycle

Space nuclear reactor power(SNRP)using a gas-cooled reactor(GCR)and a closed Brayton cycle(CBC)is the ideal choice for future high-power space missions.To investigate the safety characteristics and develop the control strategies for gas-cooled SNRP,transient models for GCR,energy conversion unit,pipes,heat exchangers,pump and heat pipe radiator are established and a system analysis code is developed in this paper.Then,analyses of several operation conditions are performed using this code.In full-power steady-state operation,the core hot spot of 1293 K occurs near the upper part of the core.If 0.4$reactivity is introduced into the core,the maximum temperature that the fuel can reach is 2059 K,which is 914 K lower than the fuel melting point.The system finally has the ability to achieve a new steady-state with a higher reactor power.When the GCR is shut down in an emergency,the residual heat of the reactor can be removed through the conduction of the core and radiation heat transfer.The results indicate that the designed GCR is inherently safe owing to its negative reactivity feedback and passive decay heat removal.This paper may provide valuable references for safety design and analysis of the gas-cooled SNRP coupled with CBC.

600-MW_(e)high-temperature gas-cooled reactor nuclear power plant HTR-PM600

The HTR-PM600 high-temperature gas-cooled reactor nuclear power plant is based on the technology of the high-temperature gas-cooled reactor pebble-bed module(HTR-PM)demonstration project.It utilizes proven HTR-PM reactor and steam generator modules with a thermal power of 250 MW_(th)and power generation of approximately 100 MW_(e)per module.Six modules in parallel,connected to a steam turbine,form a 600-MW_(e)nuclear power plant.In addition,its system configuration in the nuclear island is identical to that of the HTR-PM in which the technical risks are minimized.Under this principle,the HTR-PM600 achieves the same level of inherent safety as the HTR-PM.The concept of a ventilated lowpressure containment(VLPC)is unchanged;however,a large circular VLPC accommodating all six reactor modules is adopted rather than the previous small-cavity-type VLPC,which contains only one module,as defined for the HTR-PM.The layout of the nuclear island and its associated systems refer to single-unit pressurized water reactor(PWR)practices.With this layout,the HTR-PM600achieves a volume size of the nuclear island that is comparable to a domestic PWR of the same power level.This will be a GenerationⅣnuclear energy technology that is economically competitive.

Adaptive output-feedback power-level control for modular high temperature gas-cooled reactors

Small modular reactors(SMRs) are beneficial in providing electricity power safely and viable for specific applications such as seawater desalination and heat production. Due to its inherent safety feature, the modular high temperature gas-cooled reactor(MHTGR) is considered as one of the best candidates for SMR-based nuclear power plants. Since its dynamics presents high nonlinearity and parameter uncertainty, it is necessary to develop adaptive power-level control, which is beneficial to safe, stable, and efficient operation of MHTGR and is easy to be implemented. In this paper, based on the physically-based control design approach, an adaptive outputfeedback power-level control is proposed for MHTGRs. This control can guarantee globally bounded closedloop stability and has a simple form. Numerical simulation results show the correctness of the theoretical analysis and satisfactory regulation performance of this control.

Thermal–hydraulic analysis of space nuclear reactor TOPAZ-Ⅱ with modified RELAP5

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.

Hydrodynamic characteristics of a four-compartment periodic anaerobic baffled reactor

Periodic anaerobic baffled reactor （PABR） is a novel reactor based on the design concept of anaerobic baffled reactor （ABR）. Residence time distribution （RTD） studies on both clean and working reactors at the same hydraulic residence time （HRT） of 2 d were carded out to investigate the dead spaces and mixing patterns in PABRs at different organic loading rates （OLRs） in various switching manners and frequencies. The results showed that the fraction of dead space in PABR was similar to that in ABR, which was low in comparison with other reactor designs. Dead space may be divided into two categories, hydraulic and biological. In RTD studies without biomass, the hydraulic dead space in the PABR run in an ＂every second＂ switching manner with T = 2 d was the lowest whereas that in the PABR run in a T = ∞ （ABR） switching manner was the highest. The same trend was obtained with the total dead space in RTD studies with biomass no matter what the OLR was. Biological dead space was the major contributor to dead space but affected decreasingly at higher OLR whichever switching manner the PABR run in. The flow patterns within the PABRs were intermediate between plug-flow and perfectly mixed under all the conditions tested,

Application of Froude dynamic similitude in anaerobic baffled reactors to prediction of hydrodynamic characteristics of a prototype reactorusing a model reactor

An anaerobic baffled reactor is a system developed in recent decades and has been used as part of the treatment of high-strength wastewater. Since the function of this system is based on its hydrodynamic features, hydrodynamics and the regime of the flow through the reactor are crucial. In this study, a prototype reactor with eight chambers, which had a total volume of 48 L, and a model reactor, whose dimensions were half of those of the prototype reactor, were used. The Froude dynamic similitude in these reactors was investigated. The results show that the curve dimensionless variances were 0.089 and 0.096 for the prototype and model reactors, respectively, the short-circuiting indices were 0.483 and 0.489 for the prototype and model reactors, respectively, the effective volume and short-circuiting index measurement errors were both 1%, the hydraulic efficiency error was 2%, and the Peclet and dispersion number errors were both 7%. Most of the compared indices were close to one another in value. Therefore, the model reactor can be used based on the Froude dynamic similitude to determine hydrodynamic charac-teristics of a baffled reactor at a full scale.

Hydraulic modeling of an anaerobic expanded bed reactor for municipal sewage treatment

Anaerobic expanded bed reactor(AEBR) is mostly used for the treatment of fairly low strength wastewaters. Since the performance of AEBR largely depends on its hydraulic characteristics, residence time distribution(RTD) method is commonly used for investigation of the hydraulic characteristics of AEBR under different ascending velocity of mixed liquor. In this paper, a pilot-scale AEBR reactor is investigated for treatment of municipal sewage in which lithium chloride is used as a tracer. The results show that the AEBR could be considered as the superimposition of several constant stirred tank reactors(CSTR) and the increase of hydraulic up-flow velocity could increase the number of the CSTR and decrease the volume rate of the dead zone. The optimal up-flow velocity of the investigated AEBR was approximately 1.9 m/h in the municipal sewage treatment.

微型反应堆高速撞击下核临界安全的初步分析

微型反应堆可作为月球表面动力、航天器动力等用于太空探索,反应堆发射前需要对发射事故的影响进行安全分析。反应堆高空坠落并撞击到混凝土是主要的临界安全分析场景,高速撞击后反应堆可能重返临界。本文以两种简化的反应堆模型高速撞击场景为例,利用连续介质力学有限元程序ABAQUS与粒子输运蒙特卡罗程序耦合,完成了纯燃料堆芯垂直撞击地面和带径向反射层和屏蔽层圆柱堆30°倾角撞击地面的模拟,预测了两种场景反应堆高速撞击下keff随时间变化的物理特性。结果表明:纯燃料反应堆垂直撞击地面的keff增加最高可达1000×10^(-5),而具有反射层和屏蔽层反应堆30°倾角撞击地面的keff增加最高为200×10^(-5)。均匀密度变化条件下,采用表面非结构网格与内部非结构网格的蒙卡程序的计算结果符合较好,内部非结构网格能够更真实地捕捉材料的非均匀密度变化效应。本文的研究为微型反应堆高速撞击下的临界安全研究奠定了重要基础。

采用温度梯度的干燥反应器H_(2)O穿透深度评估方法

针对空间站CO_(2)去除系统中干燥反应器对H_(2)O穿透深度缺乏直接评估手段的问题,提出利用干燥反应器内部温度传感器的温度梯度特性来间接评估的方法;根据硅胶材料吸附H_(2)O放热、脱附H_(2)O吸热的物理特性,结合温度传感器在干燥反应器中的深度位置,分析了不同边界条件下温度传感器数据曲线在吸附及解吸周期内的变化特性,提出反映干燥反应器H_(2)O穿透深度的指标集和评估方法;并对所提方法在系统的密闭舱试验中进行了验证,验证结果表明了指标集的合理性和所提方法的有效性。

ODS MA754合金传热界面接触热阻实验研究

鉴于ODS MA754合金传热界面的接触热阻参数对全固态堆芯空间反应堆系统的热量导出具有重要影响,研发和设计了高温高压接触热阻实验装置,测量了不同温度(20~800℃)、压力(0~80 MPa)、气体氛围(He、CO_(2))以及试件表面粗糙度(1.6、3.2μm)下ODS MA754合金传热界面的接触热阻,并基于测试获得的宽量程数据点,建立了ODS MA754合金的接触热阻数据库。实验结果表明:随着接触面温度和压力的升高,界面接触热阻降低,且热阻降低的速率逐渐减小;相较于表面粗糙度为1.6μm的试件,粗糙度为3.2μm试件表面的界面接触热阻明显偏大,实验得到的定量关系可为工程样件的加工粗糙度要求提供依据;He气氛下的接触热阻远小于CO_(2)气氛,在0.1 MPa、100℃工况下,He气氛接触热阻约为CO_(2)气氛接触热阻的1/4。该研究结果可为空间反应堆的热工设计提供数据参考。

钠液面高度对氩气空间耦合传热特性影响的实验研究

锥顶盖作为快堆主容器重要的压力边界之一,其结构的完整性对快堆的安全运行至关重要。锥顶盖结构完整性与其温度载荷密切相关,而温度载荷受氩气空间自然对流与辐射耦合传热特性影响显著,有必要开展深入研究。基于钠液面上部氩气空间传热特性实验台架,开展了氩气空间传热特性实验研究,测量了锥顶盖、氩气空间和主泵支承的温度分布,阐明钠液面高度对传热特性的影响。实验结果表明,随着钠液面高度的增加,锥顶盖和氩气空间的温度随之升高,而且在高钠温时现象更加明显。此外,钠液面高度对主泵支承的温度分布和周向温差影响微弱。通过开展氩气空间钠液面高度对传热规律的影响研究,可为建立实验装置与原型反应堆之间的温度映射关系提供参考。

池式钠冷快堆复杂空间内流动与传热特性三维数值模拟研究综述

钠冷快堆是第四代先进核能系统的重要堆型之一。池式钠冷快堆安全裕量大,但结构较为复杂,其堆内一回路循环流动呈多尺度、复杂空间、多路径、三维流动等特点,给池式钠冷快堆计算与实验带来一定困难。近年来,计算流体力学(CFD)的快速发展为解决上述问题提供了重要技术路径,本文对池式钠冷快堆复杂空间内流动与传热特性三维数值模拟研究进行综述。对于全堆一体化整体计算,获取池式钠冷快堆典型对称/非对称工况下的三维温度场分布及关键热工参数瞬态变化,评价其余热排出能力;对于局部部件或区域精细化计算,获取局部三维流动与传热特征参数,也为全堆一体化计算提供关键输入。相关研究为池式钠冷快堆安全稳定运行及设计优化提供重要支撑。

热辐射特征参数对快堆锥形顶盖空间换热特性影响的研究

池式钠冷快堆的钠池内充满高温液态钠,其上方覆盖有氩气,高温液态钠主要通过辐射换热及对流换热的方式向快堆主容器上部结构及氩气空间传递热量,其换热特性及其影响因素十分复杂。辐射发射率及液态钠蒸发形成的气溶胶层对主容器上部结构的热工水力特性会造成一定影响。因此,为保证主容器上部结构在不同温度载荷下的安全性与稳定性,十分必要获取不同辐射发射率及不同气溶胶层分布下的温度分布。本研究建立大型池式钠冷快堆主容器上部锥形顶盖空间(以下简称“锥顶盖”)数值计算模型,开展数值模拟计算,得到不同发射率及气溶胶层厚度影响下锥顶盖的温度场。研究结果表明,发射率越高,锥顶盖斜肩及氩气空间温度越高;当气溶胶层存在于靠近钠液面的高度范围时(0~0.2 m),增加其厚度可使氩气空间局部温度升高,但对锥顶盖斜肩温度影响有限。发射率增高53%,则辐射换热量增加31.47%,格拉晓夫数(Gr)减少19.29%,辐射换热效果增强,自然对流效果减弱;气溶胶层高度由0增加到0~0.1 m时,对辐射换热量的吸收增加22.68%,格拉晓夫数(Gr)减少19.29%,气溶胶层高度由0~0.1 m增加到0~0.2 m时,透过气溶胶层的辐射换热量减少了0.04%,格拉晓夫数(Gr)增加了0.9%,辐射换热效果减弱,自然对流效果加强。

基于聚类和随机搜索优化的核反应堆数字孪生参数反演模型

为了实现对核反应堆内置传感器的大量数据的高效存储、传输和分析,本文结合聚类算法与随机搜索优化的人工神经网络,对空间热离子反应堆的数字孪生系统搭建了一个参数反演模型,实现在热管失效工况下的堆芯温度数据的反演。使用20%热管失效工况下空间热离子反应堆堆芯4个区域的温度数据,通过K-means聚类与轮廓系数指标提取各区域的特征温度参数,通过随机配置优化的全连接人工神经网络(ANN)完成特征参数到其他参数的反演,并对反演效果进行验证。研究结果表明,运用该方法对燃料、发射极、接收极、冷却剂4个区域进行参数反演,温度反演值的相对误差均方根分别为0.55%、0.41%、0.19%、0.18%,其中用于反演的特征参数占总参数比例均不超过8%。因此本研究建立的参数反演模型能够获取特征参数的物理含义,并对空间热离子反应堆堆芯温度参数进行较高精度的反演。

Topaz型空间堆安全棒紧急复位动力学分析

安全棒系统是Topaz型空间堆中反应性控制与核安全保护的执行机构。Topaz型空间堆在外太空轨道运行时由于堆芯温度场分布不均,反应堆内安全棒导向管因热膨胀会产生一定的径向变形。在需要紧急停堆时,安全棒紧急复位过程中受到与弯曲导向管碰撞而产生的摩擦阻力影响。本文针对导向管的径向变形量超过其与安全棒的最小间隙时,落棒时间与摩擦阻力的变化情况进行研究。基于三维多体动力学模拟软件建立了安全棒系统的动力学模型,通过精确模拟摩擦碰撞过程并合理处理刚柔耦合问题,计算分析了导向管径向变形量对安全棒紧急复位的影响。模拟结果表明,导向管径向变形量与安全棒复位时间及所受摩擦阻力均呈正相关关系。

百千瓦级空间核反应堆屏蔽优化研究

屏蔽体尺寸和重量对空间核反应堆和核动力航天器性能有着重要影响,因而屏蔽设计优化是空间核动力系统设计的关键。本文以JIMO项目反应堆为对象,在铍-碳化硼-钨-氢化锂分层组合屏蔽方案的基础上,考虑到辐照剂量的径向分布,采用蒙特卡罗方法计算了负载处辐照剂量和氢化锂中子剂量,分析了屏蔽设计原理,并提出了分步优化方法以实现屏蔽优化。根据结果分析,调整了铍和碳化硼的厚度比例、钨半径及布置位置,获得了优化的屏蔽方案,在满足屏蔽要求的基础上质量减少了98.41 kg。提出的屏蔽方案及设计流程可为空间核电源屏蔽设计优化提供参考。

空间热管反应堆电源研究进展及展望

深空探测技术的发展对动力系统提出了更高的要求。传统的太阳能电源与化学电源的适用范围较小,环境适应能力不强,而微型核反应堆电源能量密度高,不依赖太阳光照,可应用于轨道运输、高轨探测多场景任务。在微型核反应堆电源技术路线中,热管冷却核反应堆电源因其系统设备极大简化、模块化设计,高可靠的全固态堆芯、非能动传热及瞬态响应迅速等特性,成为空间核反应堆电源最具可行性的路线之一。通过文献调研总结目前空间热管堆发展现状,从发展历史出发,梳理热管冷却核反应堆电源设计和理论研究,总结热管冷却核反应堆电源发展方向和关键技术。

空间热离子反应堆燃料元件力学性能分析程序开发

为了实现对空间热离子反应堆燃料元件运行期间安全性能的预测,研究开发了一种燃料元件力学性能分析程序,并针对多层圆筒状的TOPAZ-Ⅱ热离子燃料元件开展了应力、应变和几何变形的高精度模拟。程序考虑了核燃料在高温辐照环境下的辐照肿胀,并分析了燃料芯块-发射极在发生接触后的力学响应问题,从而快速且准确地求解燃料芯块和发射极的力学状态,以对空间热离子反应堆运行期间的性能提供准确预测。结果表明:在正常运行情况下,空间热离子反应堆燃料会发生显著的肿胀效应,其造成的变形将导致燃料元件热电转换效率降低、元件失效等安全隐患。

空间堆线性变参数模型预测控制方法

针对传统PID控制方法在面临复杂负荷需求变化时无法满足空间核反应堆的控制性能需求的问题,提出了基于线性变参数模型的空间堆模型预测控制方法。针对强非线性特性,采用线性变参数模型作为预测模型,对全功率范围内的空间堆运行特性进行预测;针对大时滞的特性,采用串级控制理论结合模型预测方法优化执行机构动作,提升被控量的响应速度。分别设计了电功率和冷却剂温度的预测控制系统,在典型工况下进行控制性能仿真验证,结果表明:相对于串级PID控制系统,所设计的预测控制系统的电功率调节时间减少了68.7%,冷却剂温度调节时间减少了81.7%;电功率和冷却剂温度预测控制系统的核功率的超调量分别减少了22.3%和13.0%,保证了反应堆的安全;控制系统的适用性强,在不同功率水平均具有良好的控制性能;执行机构的动作大幅减少,驱动电机峰值电压下降了84.1%,使用寿命得到延长,系统经济性得到提升。该研究提出的结合串级控制理论基于线性变参数模型的预测控制方法能够用于优化空间堆的控制系统。