模块式高温气冷堆的固有安全特性

通过对高温气冷堆安全特性的研究,简要分析了高温气冷堆阻止放射性释放的多重屏障、反应性瞬变的固有安全性、非能动的余热排出系统及其他安全特性,从而表明高温气冷堆具有固有安全性的特点。     

10MW高温气冷实验堆电力系统

10MW高温气冷堆电力系统是为保障反应堆完成其安全目标和运行目标而设置的重要系统,本文介绍了10MW高温气冷堆电力系统的组成结构和配置,描述了电力系统在不同工况下的运行方式。     

高温气冷堆产业推广及应用前景

高温气冷堆是我国具有完全自主知识产权的第四代先进核能技术,具有固有安全性、模块化设计及建造、发电效率高、用途广泛等特点。文章介绍了高温气冷堆产业化推广以及高温气冷堆在替代中小火电、制氢、石化和海水淡化等领域多功能综合利用的发展前景,分析了高温气冷堆产业化面临的挑战,指明了高温气冷堆产业化发展的重要意义。     

10MW高温气冷实验堆乏燃料元件贮存中放射性裂变产物的安全分析

高温气冷堆乏燃料元件的放射性裂变产物绝大部分滞留在燃料元件中。10MW高温气冷实验堆在设计寿命内将卸出约9万个乏燃料元件,其放射性裂变产物的活度高达1.9×1017Bq,因此正确实施乏燃料元件的贮存,减少放射性裂变产物向环境中释放和进行有效的屏蔽是极其重要的。本文根据乏燃料元件中放射性裂变产物的计算结果和德国高温气冷堆乏燃料元件贮存的经验．对我国10MW高温气冷堆乏燃料元件贮存中放射性裂变产物进行了安全分析。     

10MW高温气冷堆的动态特点与运行程序模型

介绍了10MW高温气冷堆系统结构及特点，给出了几个物理量变化所引起的其它物理量变化的模拟计算结果。根据系统结构特点以及模拟计算做出了10MW高温气冷堆总的运行程序模型图以及在几个运行阶段的调控方法。     

HTR-10应急电力系统设计及其调试

在充分利用高温气冷堆所具有的良好固有安全性这一特点的基础上，以安全级不间断电源装置组成了静止式新型应急电源系统，从而大大地减化了系统，降低了投资，并显著地提高了系统的安全性和可靠性，本文全成介绍了10MW高温气冷实验堆（HTR－10）应急电力系统的系统功能，主要设计原则，系统组成以及应急电力系统的核心设备－安全级不间断电源装置的运行方式和安全特性，并系统地总结了应急电力系统的调试工作。     

小型高温气冷堆的堆型方案比较

本文对模块式高温气冷堆的棱柱状和球床两种堆芯型式和一体化与肩并肩分置式两种总体设计方案分别进行了技术特点、设计制造、运行经验和安全性与经济性的比较,提出了在我国发展高温气冷堆的堆型选用原则和建议.     

模块式高温气冷堆的技术背景及展望

模块式高温气冷堆具有固有安全性、发电效率高、用途广泛等特点,是第四代核能系统代表堆型之一,也是我国16个重大科技专项之一。本文介绍了高温气冷堆的发展历史,对高温气冷堆国际研究现状进行了阐述,说明了高温气冷堆在我国的发展情况。介绍了我国正处于调试期的模块式高温气冷堆示范电站的技术特点,从高效发电、工艺热应用、能源替代、分布式能源四个角度对模块式高温气冷堆的发展前景进行了分析,提出了我国模块式高温气冷堆后续工作建议。     

球床模块式高温气冷堆失冷事故特性研究

利用高温气冷堆专用系统分析软件THERMIX程序,对球床模块式高温气冷堆(HTR-PM)失冷失压和失冷不失压事故的动态特性进行了研究,分析了堆芯功率、燃料最高温度及堆舱水冷壁余热载出功率等关键参数的变化过程,并对影响余热排出功率和燃料最高温度的不确定性进行了评价.研究结果表明,在失冷事故下,堆芯余热可通过热传导、辐射和自然对流等非能动方式传至最终热阱大气,燃料元件和压力容器等重要部件的最高温度均在设计限值内.这为HTR-PM保持模块式高温气冷堆固有安全性不变的同时实现单堆250 MW的功率方案奠定了基础,也为后续高温气冷堆电站示范工程进一步的深入设计研究提供了依据.     

HTR-10安全负压通风系统的设计

由于高温气冷堆具有固有安全性，因此在高温气冷堆的设计中采用通风式包容体替代了密封承压式安全壳，在供暖，通风与空调（HVAC）系统中相应采用了安全负压通风系统，以保证包容体在正常工况或事故工况下都能满足与安全相关的一切功能，本文介绍了10MW高温气冷实验堆（HTR－10）安全负压通风系统的设计与评价。     

球床模块式高温气冷堆核电站的概率安全分析框架

通过对国际上相似堆型概率安全分析(PSA)框架的调研,结合球床模块式高温气冷堆(HTR-PM)自身设计特点,提出以始发事件为起点,以事件序列为主干,以释放类为终点的HTR-PM的PSA一体化事件树框架.分析表明,HTR-PM在PSA框架上的特点主要由其设计特点决定.     

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).     

Reactor core design of Gas Turbine High Temperature Reactor 300

Japan Atomic Energy Research Institute (JAERI) has been designing Japan’s original gas turbine high temperature reactor, Gas Turbine High Temperature Reactor 300 (GTHTR300). The greatly simplified design based on salient features of the High Temperature Gas-cooled Reactor (HTGR) with a closed helium gas turbine enables the GTHTR300 a highly efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the High Temperature Engineering Test Reactor (HTTR) and existing fossil fired gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original design features of this system are the reactor core design based on a newly proposed refueling scheme named sandwich shuffling, conventional steel material usage for a reactor pressure vessel (RPV), an innovative coolant flow scheme and a horizontally installed gas turbine unit. The GTHTR300 can be continuously operated without the refueling for 2 years. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200,000 yen (1667 US$)/kW e, and the electric generation cost is close to a target cost of 4 yen (3.3 US cents)/kW h.

This paper describes the original design features focusing on the reactor core design and the in-core structure design, including the innovative coolant flow scheme for cooling the RPV. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan.     

The design features of the HTR-10

The 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is a modular pebble bed type reactor. This paper briefly introduces the main design features and safety concept of the HTR-10. The design features of the pebble bed reactor core, the pressure boundary of the primary circuit, the decay heat removal system and the two independent reactor shutdown systems and the barrier of confinement are described in this paper.     

Achievement of Reactor-Outlet Coolant Temperature of 950°C in HTTR

A High Temperature Gas-cooled Reactor (HTGR) is particularly attractive due to its capability of producing high-temperature helium gas and to its inherent safety characteristics. The High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan, achieved its rated thermal power of 30 MW and reactor-outlet coolant temperature of 950°C on 19 April 2004. During the high-temperature test operation which is the final phase of the rise-to-power tests, reactor characteristics and reactor performance were confirmed, and reactor operations were monitored to demonstrate the safety and stability of operation. The reactor-outlet coolant temperature of 950°C makes it possible to extend high-temperature gas-cooled reactor use beyond the field of electric power. Also, highly effective power generation with a high-temperature gas turbine becomes possible, as does hydrogen production from water. The achievement of 950°C will be a major contribution to the actualization of producing hydrogen from water using the high-temperature gas-cooled reactors. This report describes the results of the high-temperature test operation of the HTTR.     

LOCA Analysis of Super Fast Reactor

This paper describes loss of coolant accident (LOCA) analyses of the Supercritical-pressure Water-Cooled Fast Reactor (Super Fast Reactor). The features of the Super Fast Reactor are high power density and downward flow cooled fuel channels for the improvement of the economic potential of the Super Fast Reactor with high outlet steam temperature. The LOCA induces large pressure and coolant density change in the core. This change influences the flow distribution among the downward flow parallel channels. It will affect the safety of the Super Fast Reactor. LOCA analysis of Super Fast Reactor is important to understand the safety features of the Super Fast Reactor. Keeping the flow rate in the core is important for the safety of the Super Fast Reactor. In LOCA, it is difficult to maintain an adequate flow rate due to the once-through coolant cycle and the downward flow cooled fuel assemblies. Therefore, the early actuation of the Automatic Depressurization System (ADS) and reduction of the maximum linear heat generation rates of the downward flow seed fuel assemblies and Low-Pressure Core Spray (LPCS) system are necessary for the Super Fast Reactor to cool the core under LOCA. Analysis results show that the Super Fast Reactor can satisfy the safety criteria with these systems.     

The reactor dynamics code DYN3D – models,validation and applications

The article provides an overview of the reactor dynamics code DYN3D. The code comprises various 3D neutron kinetics solvers, a thermal-hydraulics reactor core model and a thermo-mechanical fuel rod model. The implemented models and methods and the capabilities and features of the code are described. Latest developments of models and methods are delineated. An overview on the status of verification and validation is given. Code applications for selected safety analyses are described. Furthermore, multi-physics code couplings to thermal-hydraulic system codes, CFD and sub-channel codes as well as to the fuel performance code TRANSURANUS are outlined. Developments for innovative reactor concepts, in particular Molten Salt Reactor, High Temperature Gas-cooled Reactor and Sodium Fast Reactor are delineated. The management of code maintenance is briefly described. An outlook on further code development is given.     

高温气冷堆蒸汽发生器换热管断裂事故进水量影响因素分析

以清华大学核能与新能源技术研究院设计的250 MW球床模块式高温气冷堆(HTR-PM)为例,对蒸汽发生器换热管断裂事故下影响一回路进水量的一些因素进行了分析.分析结果表明:除了断管位置、破口面积等对一回路进水量有直接影响外,进水量还与泄放管线直径、节流孔直径、泄放阀门选择、泄放系统动作设定等因素有关.合理地选择参数可有效排空蒸汽发生器内存留的水,避免一回路大量进水并减少一回路放射性物质向二次侧泄漏所造成的污染.     

200MW池式供热堆未能紧急停堆的预期瞬变事故分析

利用轻水堆系统通用的热工水力分析程序ＴＥＴＲＡＮ－０２,对２００ＭＷ池式供热堆的未能紧急停堆的预期瞬变事故。即断电ＡＴＷＳ事故,误提棒ＡＴＷＳ事故,外负荷丧失ＡＴＷＳ事故等进行了计算和分析。结果表明,在事故过程中,订参数没有超出鸡范围;不需任何设备动作和人员干预,反应堆就能自动降功率,维持长期堆芯冷却,具有较高的安全性。