The Shandong Shidao Bay 200 MW_e High-Temperature Gas-Cooled Reactor Pebble-Bed Module(HTR-PM) Demonstration Power Plant: An Engineering and Technological Innovation

In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor （HTGR） by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current status is that high temperature reactors have been relegated to completing research programs on advanced fuels, graphite and materials with no plans to build a demonstration plant as required by the US Con- gress in 2005. There are many reasons behind this diminution of HTGR development, including but not limited to insufficient government funding requirements for research, unrealistically high temperature requirements for the reactor, the delay in the need for a ＂hydrogen＂ economy, competition from light water small modular light water reactors, little utility interest in new technologies, very low natural gas prices in the US, and a challenging licensing process in the US for non-water reactors.

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.

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%.

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 of10MW High Temperature Reactor Test Modulegh

he thermal performance test of the horizontal coaxial double tube hot gas duct (HGD) with an internal thermal insulation for the 10MW High Temperature Reactor Test Module (HTR10) was conducted on a Helium Test Loop(HETL). The present paper deals with the technical feature of the HETL, the test section and the thermal performance test of the HGD. The HGD test section with a triple tube structure includes an inner heater, a HGD model and a coldhot gas mixer. A counterflow of cold and hot helium gas under the pressure of about 3.0 MPa and the minimum temperature of 100℃ in the annular passage and the maximum of 950℃ in the central tube of the HGD model was formed. The HGD model was undergone 20 times of pressure cycle test under the pressure ranging from 0.1 to 3.4 MPa, 18 times of the temperature cycle test under the temperature ranging from 100 to 950℃ and high temperature (700 to 950℃) helium flow test for a period of more than 350 hours. The effective thermal conductivity (λeff) of the internal insulation of the HGD was investigated experimentally. The relationship of the effective thermal conductivity with the average tmperature of the internal insulation layer is λeff(W/m/℃)=0.3512+0.0003T(℃). The test results indicate that the HGD model has good abilities to resist heat flux from the central tube to the annular passage, temperature variations, and pressure variations.

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.

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.

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.

Self-acting Afterheat Removal in High Temperature Gas Cooled Reactors

Much more nuclear energy capacity is needed than currently installed to meet the demand of energy and the requirement on environment protection in the next decades. More stringent nuclear safety standards have to be established for future nuclear power plants.The philosophy of a catastrophe free nuclear technology is presented in this paper. The issue of afterheat removal of high temperature gas cooled reactors is handled.It is a striking inherent safety feature of the modular high temperature gas cooled reactor design that the afterheat removal takes place without any active core cooling systems.

HTR-PM汽轮机轴封蒸汽断供原因分析及解决方案

介绍了高温气冷堆示范工程(HTR-PM)汽轮机轴封系统,并且对轴封蒸汽断供的原因进行了分析,提出了可以采用蒸汽发生器及管道内余汽供轴封、辅助电锅炉至轴封供汽管道改造、高温空气供轴封这3种解决方案。对3种方案进行对比,分析其在成本、可靠性和运行限制方面的优缺点,结果表明3种方案均是可行的。研究结果为HTR-PM汽轮机轴封蒸汽断供问题的解决提供了参考。

Monte Carlo studies on the burnup measurement for the high temperature gas cooling reactor

Online fuel pebble burnup measurement in a future high temperature gas cooling reactor is proposed for implementation through a high purity germanimn （HPGe） gamma spectrometer. By using KORIGEN software and MCNP Monte Carlo simulations, the single pebble gamma radiations to be recorded in the detector are simulated under different, irradiation histories. A specially developed algorithm is applied to analyze the generated spectra to reconstruct the gamma activity of the ~arCs monitoring nuclide. It is demonstrated that by taking into account the intense interfering peaks, the 137Cs activity in the spent pebbles can be derived with a standard deviation of 3.0% （l（r）. The results support the feasibility of utilizing the HPGe spectrometry in the online determination of the pebble burimp in future modular pebble bed reactors.

Characteristics and accumulation mechanisms of the Dongfang 13-1 high temperature and overpressured gas field in the Yinggehai Basin, the South China Sea

The Dongfang 13-1 is located in the diapiric structure belt of the Yinggehai Basin. The formation pressure of its main gas res- ervoir in the Miocene Huangliu Formation is up to 54.6 MPa （pressure coefficient=l.91） and the temperature is as high as 143°C （geothermal gradient 4.36°C/100 m）, indicating that it is a typical high-temperature and overpressured gas reservoir. The natural gas is interpreted to be coal-type gas derived from the Miocene mature source rocks containing type Ⅱ2-Ⅲ kero- gens as evidenced by high dryness index of up to 0.98 and heavy carbon isotopes, i.e., the δ13C2 ranging from -30.76%o to -37.52%o and δ13C2 ranging from -25.02%o to -25.62%o. The high temperature and overpressured Miocene petroleum system is related mainly to diapir in the Yinggehai Basin and contains more pore water in the overpressured reservoirs due to under- compaction process. The experimental and calculated results show that the solubility of natural gas in formation water is as high as 10.5 m3/m3 under the temperature and pressure conditions of the Sanya Formation, indicating that at least part of the gas may migrate in the form of water-soluble phase. Meanwhile, the abundant gas source in the Basin makes it possible for the rapid saturation of natural gas in formation water and exsolution of soluble gas. Therefore, the main elements controlling formation of the Dongfang 13-1 gas pool include that （1） the diapir activities and accompanying changes in temperature and pressure accelerate the water-soluble gas exsolution and release a lot of free gas; （2） submarine fan fine sandstone in the Huangliu Formation provides good gas-water segregation and accumulation space; and （3） the overlying overpressured mud rocks act as effective caps. The accumulation mechanism reveals that the high temperatural and high pressure structure belt near the diapir structures has a good potential for large and medium-sized gas field exploration.

HTR-10燃料元件装卸系统冷态调试

10MW高温气冷实验堆(HTR-10)是一座球床堆,由燃料元件装卸系统实现燃料元件的装卸和循环,且不需要停堆。为保证HTR-10的正常运行,燃料元件装卸系统必须安全、可靠。为此,必须对燃料元件装卸系统进行周密、细致的调试试验和验证。本文介绍了燃料元件装卸系统冷态调试的主要调试项目、调试方法和调试结果。

溶胶-凝胶法制备二氧化铀核芯的U_3O_8欠酸溶解工艺

在UO2核芯的制备工艺中,为获得高浓度铀的硝酸铀酰溶液,同时降低溶液中硝酸根含量,即获得低于硝酸铀酰标准化学计量比(硝酸铀酰中硝酸根与铀酰离子化学计量比为2)的溶液,必须采用欠酸溶解工艺。本文采用分批加料和阶段性加热方式获得了很好的溶解效果,得到的硝酸铀酰溶液中U含量为2.1～2.5mol/L,NO3-和U的摩尔浓度比为1.6～1.8,溶液pH大于1.4,从而成功制备出合格的欠酸溶解的硝酸铀酰溶液。在此基础上,根据多次溶解试验的结果总结出了溶液密度与U含量的经验公式。

HTR-10蒸汽发生器设计

描述了HTR10 蒸汽发生器的结构设计, 热工水力学特性, 材料选择和力学分析等。作为反应堆一回路和二回路之间进行热交换的关键设备, HTR10 蒸汽发生器采用了独特的管束设计,

HTR-10主氦风机性能试验研究

主氦风机是 10MW高温气冷实验堆 (简称HTR 10 )的关键设备 ,其功能是提供一回路氦气循环所需的压力升 ,从而使堆芯的热量传输给蒸发器。主氦风机的工作条件是 2 5 0℃ ,3 0MPa的氦气。主氦风机的性能试验在制造厂进行 ,为了降低试验难度采用氮气模拟试验方案。本文对主氦风机性能试验的原理和过程进行了描述 ,对试验结果进行了分析。此外 ,还对主氦风机在HTR 10上的运行特性进行了预测分析。结果表明 ,主氦风机气动性能完全能满足在HTR 10上的运行需要 ,并且有很大裕量。

LBB思想在HTR-10蒸汽发生器传热管上的应用

试图将LBB思想应用到高温气冷堆蒸汽发生器传热管上 ,按其中的若干原则规定 ,对蒸汽发生器传热管进行了LBB分析 ,初步结果表明传热管满是LBB和件 ,具有较大的安全裕度。对SG管束进行的LBB分析初步结果显示SG管束满足LBB准则。

HTR-PM直流蒸汽发生器的建模与分析

球床模块式高温气冷堆核电站示范工程(HTR-PM)采用两座模块式高温气冷堆带一台汽轮发电机组的技术方案,为了开展其运行特性研究,清华大学核能与新能源技术研究院开发了针对HTR-PM的工程模拟机,其中螺旋管式直流蒸汽发生器的模型还需进一步完善。本文深入分析了螺旋管式直流蒸汽发生器的流动、换热规律,明确了蒸汽发生器一次侧和二次侧的流动与换热模型,通过对稳态工况中分布数据的详细分析,说明了模拟结果的正确性。为适应更多模块的高温气冷堆核电站的运行分析要求,通过网格划分方案的讨论与优化,在保证实时性的前提下,提高了蒸汽发生器中流动与换热模拟的准确性,为下一步采用工程模拟机开展其运行特性研究打下基础。