Design of the material performance test apparatus for high temperature gas-cooled reactor

Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor(HTGR).To solve the problem,a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the HTGR.The apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the apparatus.Specially designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field.A typical experimental process was analyzed in the paper,which lasted 76 hours including seven stages.Experimental results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and verified.Test temperature in the apparatus could be elevated up to 1600oC,which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor.

Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor

The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calcu- lated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thermophore- sis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen’s formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell’s temperature. Its maximum value is 14%.

Safety Features of Modular High Temperature Gas-cooled Reactors (MHTGR)

The following design features which satisfy fundamental safety design objectives of an MHTGR are analyzed: (i) inherent safety features to reactivity effect: (ii) passive decay heat removal: and (iii) multiple barriers.Several events have been identified to be the bounding. hypothetical accidents for the MHTGR. The important accident sequences leading to severe accidents are ingress of a large amount of water or air into the core. The analyses of severe accident scenarios have shown that even the harm of fuel element predicted to occur by chmeical reaction after a hypothetical large amount of water ingress into the core or air ingress into the core will not result in major impact on the environment due to the nitegrity of fuel particles remained. Therefore, it would not be necessary to require an emergency plan to evacuate nearby inhabitants.

Numerical investigations of thermal mixing performance of a hot gas mixing structure in high-temperature gas-cooled reactor

A numerical simulation study was performed to clarify the thermal mixing characteristics of coolant in the core bottom structure of the high-temperature gas-cooled reactor(HTR). The flow field and temperature field in the hot gas chamber and the hot gas duct of the HTR were obtained based on the commercial computational fluid dynamics(CFD) program. The numerical simulation results showed that the helium flow with different temperatures in the hot gas mixing chamber and the hot gas duct mixed intensively, and the mixing rate of the temperature in the outlet of the hot gas duct reached 98 %. The results indicated many large-scale swirling flow structures and strong turbulence in the hot gas mixing chamber and the entrance of the hot gas duct, which were responsible for the excellent thermal mixing of the hot gas chamber and the hot gas duct. The calculated results showed that the temperature mixing rate of the hot gas chamber decreased only marginally with increasing Reynolds number.

Study on neutronics design of ordered-pebble-bed fluoride-salt- cooled high-temperature experimental reactor

This paper presents a neutronics design of a 10 MW ordered-pebble-bed fluoride-salt-cooled high-temperature experimental reactor. Through delicate layout, a core with ordered arranged pebble bed can be formed,which can keep core stability and meet the space requirements for thermal hydraulics and neutronics measurements.Overall, objectives of the core include inherent safety and sufficient excess reactivity providing 120 effective full power days for experiments. Considering the requirements above, the reactive control system is designed to consist of 16 control rods distributed in the graphite reflector. Combining the large control rods worth about 18000–20000 pcm, molten salt drain supplementary means(-6980 to -3651 pcm) and negative temperature coefficient(-6.32 to -3.80 pcm/K) feedback of the whole core, the reactor can realize sufficient shutdown margin and safety under steady state. Besides, some main physical properties, such as reactivity control, neutron spectrum and flux, power density distribution, and reactivity coefficient,have been calculated and analyzed in this study. In addition, some special problems in molten salt coolant are also considered, including ~6Li depletion and tritium production.

Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment

Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized water reactor(PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gascooled reactor(HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage(hundreds of degrees centigrade). The tristructuralisotropic(TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level3 PSA is done.

Thermal Performance Test of the Hot Gas Duct of 10MW High Temperature Reactor Test Module

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Gearbox Scheme in High Temperature Reactor Helium Gas Turbine System

Helium Turbine is used in High Temperature Reactor Helium Gas Turbine (HTR-GT) system, by which the direct helium circulation between the reactor and turbine generator system will come true. Between helium turbine and generator, there is gearbox device which reduces the turbine rotation speed to normal speed required by the generator. Three optional gearbox schemes are discussed. The first is single reduction cylindrical gearbox, which consists of one high speed gear and one low speed gear. Its advantage is simple structure, easy to manufacture, and high reliability, while its disadvantage is large volume and misalignment of input and output axle. The second is planetary gear mechanism with static planet carrier. The third is planetary gear mechanism with static internal gear. The latter two gearbox devices have similar structure. Their advantage is small volume and high reduction gear ratio, while disadvantage are complicated structure, many gears, low reliability and low mechanical efficiency.

The AVR as Small Modular Thorium Very High Temperature Reactor:Experiences-Design-Safety-Fuel Cycle

As a young engineer in the power plant department of Brown Boveri,Dr.Schulten had the idea to design nuclear power stations without major risk.The following requirements must be accomplished:ŸA negative temperature coefficient had to avoid an MCA(Maximum Credible Accident);ŸCeramic materials for core construction and fuel elements;ŸA homogenous mixture of nuclear fuel and graphite had to be able to use uranium and thorium as breeding material;ŸThe produced high temperature heat shall be the basis for production of electricity,drinking water,hydrogen,etc.;ŸA relatively simple plant,which could be operated in developing countries,to cogenerate electricity and heat;ŸHelium used as cooling gas.

Conceptual design and safety characteristics of a new multi-mission high flux research reactor

Research reactors with neutron fluxes higher than 10^(14) n cm^(−2) s^(−1) are widely used in nuclear fuel and material irradiation,neutron-based scientific research,and medical and industrial isotope production.Such high flux research reactors are not only important scientific research facilities for the development of nuclear energy but also represent the national comprehensive technical capability.China has several high flux research reactors that do not satisfy the requirements of nuclear energy development.A high flux research reactor has the following features:a compact core arrangement,high power density,plate-type fuel elements,a short refueling cycle,and high coolant velocity in the core.These characteristics make it difficult to simultaneously realize high neutron flux and optimal safety margin.A new multi-mission high flux research reactor was designed by the Institute of Nuclear and New Energy Technology at Tsinghua University in China;the reactor can simul-taneously realize an average neutron flux higher than 2.0×10^(15) n cm^(−2) s^(−1) and fulfill the current safety criterion.This high flux research reactor features advanced design concepts and has sufficient safety margins according to the preliminary safety analysis.Based on the analysis of the station blackout accident,loss of coolant accident,and reactivity accident of a single-control drum rotating out accidently,the maximum temperature of the cladding surface,minimum departure from nucleate boiling ratio,and temperature difference to the onset of nucleate boiling temperature satisfy the design limits.

Macroscopic Structural Analysis on a 10 kW Class Lab-Scale Process Heat Exchanger Prototype under a High-Temperature Gas Loop Condition

A PHE (Process Heat Exchanger) is a key component in transferring high-temperature heat generated from a VHTR (Very High Temperature Reactor) to a chemical reaction for the massive production of hydrogen. Last year, a 10 kW class lab-scale PHE prototype made of Hastelloy-X was manufactured at the Korea Atomic Energy Research Institute (KAERI), and a performance test of the PHE prototype is currently underway in a small-scale nitrogen gas loop at KAERI. The PHE prototype is composed of two kinds of flow plates: grooves 1.0 mm in diameter machined into the flow plate for the primary coolant, and waved channels bent into the flow plate for the secondary coolant. Inside the 10 kW class lab-scale PHE prototype, twenty flow plates for the primary and secondary coolants are stacked in turn. In this study, to understand the macroscopic structural behavior of the PHE prototype under the steady-state operating condition of the gas loop, high-temperature structural analyses on the 10 kW class lab-scale PHE prototype were performed for two extreme cases: in the event of contacting the flow plates together, and when not contacting them. The analysis results for the extreme cases were also compared.

Simultaneous approach for simulation of a high-temperature gas-cooled reactor

The simulation of a high-temperature gas-cooled reactor pebble-bed module(HTR-PM) plant is discussed.This lumped parameter model has the form of a set differential algebraic equations(DAEs) that include stiff equations to model point neutron kinetics.The nested approach is the most common method to solve DAE,but this approach is very expensive and time-consuming due to inner iterations.This paper deals with an alternative approach in which a simultaneous solution method is used.The DAEs are discretized over a time horizon using collocation on finite elements,and Radau collocation points are applied.The resulting nonlinear algebraic equations can be solved by existing solvers.The discrete algorithm is discussed in detail;both accuracy and stability issues are considered.Finally,the simulation results are presented to validate the efficiency and accuracy of the simultaneous approach that takes much less time than the nested one.

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。

高温气冷堆热态功能试验中一回路加热动态特性研究

高温气冷堆反应堆压力容器堆内构件由石墨和碳化硼烧结陶瓷材料组成,热态功能试验过程中需对一回路堆内构件进行加热除湿。针对高温气冷堆一回路热试系统特点,采用COMSOL Multiphysics 5.4计算软件构建了压力容器、蒸汽发生器和主氦风机三维数值模型,并通过示范工程加热试验值验证了模型的可靠性,并获得了一回路加热过程中温度场动态特性。为了解决高温气冷堆示范工程一回路首次加热效率低的问题,提出了将辅助蒸汽通入到蒸汽发生器,并动态调整辅助蒸汽运行参数来加热一回路氦气的方法,结果表明:相较于主氦风机单独加热方式,外加辅助蒸汽热源可节省加热时间约31.3 h,同时可将堆芯最终平衡温度由250℃提高至265℃;在满足运行准则的前提下,更有利于碳砖和石墨堆内构件的除湿。该研究结果为高温气冷堆一回路热试提供了有力支持。

清华大学核研院研制5 MW低温核供热试验堆与10 MW高温气冷实验堆的工程技术创新

清华大学核能与新能源技术研究院(简称核研院)先后在1989年和2000年建成了5 MW低温核供热试验堆与10 MW高温气冷实验堆。在建堆过程中,清华大学核研院坚持设计创新与工具创新、工艺创新、工序创新密切结合,完成了一系列关键设备和零部件的制造与安装,使得整个工程项目顺利完工。在工程史研究中,技术工人做出的创新贡献并未引起学术界足够重视。本文表明,技术工人在工具、工艺、工序、制造与安装阶段的技术创新,亦是工程创新的重要保证。

高温气冷堆主氦风机变阻力工况调试方法

高温气冷堆主氦风机调试期间,由于一回路阻力低于设计工况,因而主氦风机无法完成全转速范围性能测试。基于主氦风机的理论特性与相似原理开发主氦风机不同阻力工况调试参数的推算方法。结合主氦风机单体试验工况点,准确推算主氦风机冷态与热态性能试验的工况点参数,并指导完成高温气冷堆主氦风机全转速、满功率性能测试。通过对主氦风机调试与出厂试验结果的对比分析,验证本推算方法的可行性,并给出空气介质与氦气介质工况转换的修正因子。通过主氦风机调试与运行数据的对比分析,可以看出本文提供的主氦风机调试工况具有足够的包络性,能够覆盖高温气冷堆运行期间主氦风机的所有运行工况,证明了本文提供的变阻力工况主氦风机调试方法满足高温气冷堆主氦风机性能验证需求,可用于指导后续高温气冷堆的主氦风机调试工作。

氦气风机驱动电机转子径向通风结构研究

氦气风机驱动电机是高温气冷堆一回路唯一能动设备,其安全性直接影响到高温气冷堆的安全稳定运行。针对高温高压环境及氦气传热工质对驱动电机风路结构设计及相关参数影响引发的新问题,通过建立双侧转子铁心三维风路结构流固耦合传热模型,计算分析了高温高压环境下传热工质氦气在转子径向风沟内的流动特性,研究了转子分别在动态和静态情况下,通风结构内流体流动规律及转子铁心和导条的温升变化规律。同时,对不同种类传热工质的对应规律进行了对比研究,搭建了模拟实验测试装置,实测数据与数值计算结果吻合较好,验证了计算分析结果的合理性和正确性,为氦气风机驱动电机转子风路结构优化、冷却性能的有效提升提供一定的理论参考。

双碳目标下高温气冷堆替代中小型火电的思考

碳达峰碳中和是我国一项重要的国家战略决策,“双碳”目标的提出给能源行业带来了深刻的变革,我国能源结构将进一步优化,电力行业必须实现低碳转型,新能源取代火电已成为必然。本文分析了新形势下火电企业面临的巨大挑战、发电规模受到限制、经营成本不断上升,尤其是中小型火电机组,面临着淘汰和关停的局面,而高温气冷堆具有固有安全性、发电效率高、用途广泛的显著特点,使其替代中小型火电成为可能。本文通过厂址适应性、技术可行性、经济可行性三方面,分析了高温气冷堆替代中小型火电的可行性,并从相关法律法规适用性、内陆厂址所带来的问题、开展公众沟通等方面提出建议。

紧凑新型高温超导可控电抗器的仿真研究

为解决电网配网侧无功就地消纳的需求,文章提出了一种紧凑新型高温超导可控电抗器,介绍了该紧凑新型高温超导可控电抗器工作原理与基本结构,搭建了有限元分析模型并得到其电感调节特性,以期为相关人员提供参考。