模块式高温气冷堆核能系统二回路流体网络控制特性分析

针对模块式高温气冷堆（MHTGR）核能系统二回路流体网络进行非线性建模,研究管路动力学特性及网络拓扑结构特性,建立了微分-代数模型,设计了模块质量流量和汽机主蒸汽压力的调控方案。在MATLAB/Simulink环境下对模型进行标准化封装,以高温气冷堆核电站示范工程（HTR-PM）为例进行二回路流体网络的仿真。结果表明,模型有效地反映了系统二回路流体网络的非线性特性,设计的控制器使得模块流体质量流量和汽机主蒸汽压力有效地收敛于参考值,各项控制指标均高于控制要求。设计的仿真平台可为实际工程调控中积分时间系数的选择、拥有更多模块数量的高温气冷堆核能系统二回路流体网络的调控等提供试验仿真测试。     

高温气冷堆二回路加药系统工艺优化及自动控制方法研究

目前高温气冷堆（HTGR）的二回路水质指标无相关的成熟标准可以参考，加药控制指标也没有标准。本文通过对HTGR二回路的材质和运行工况的研究，参考压水堆（PWR）和火电直流炉的运行经验，确定二回路给水pH值控制标准在9.5~9.8之间，联氨的控制标准在80~120 μg/L之间可以使二回路取得较好的防腐效果；针对HTGR二回路加药系统设计现状，对加药系统设计工艺提出了优化和变更方案，用联氨表代替溶氧表，用电导率计算pH，避免了溶氧表和pH表的滞后性和不稳定性；最后，通过控制方法的改进，实现HTGR二回路加氨和加联氨系统的全自动精准控制。      

10MW高温气冷堆控制方案研究

在清华大学核能设计研究院开发的高温堆可视化仿真控制平台上进行了10MW高温气冷堆动态特性研究,并结合其运行特点和控制要求设计了3种控制方案,采用比例积分与微分控制方法,在高温堆可视化仿真控制平台上进行了控制方案的仿真比较。控制的重点在于维持直流蒸汽发生器的出口蒸汽温度恒定,同时兼顾反应堆出口热氦气温度不超出保护限值。仿真结果表明,采用给水泵调节给水流量来控制蒸汽温度,并通过氦风机调节氦流量保持与给定功率成比例,避免跨回路调节,静态解除了由于氦流量的变化对一、二回路的耦合问题,能够获得理想的控制效果。     

采用虚拟场景技术的高温气冷堆仿真系统研究

仿真系统对10 MW高温气冷堆的堆芯、主回路系统和蒸汽发生器等部件进行分析计算,模拟稳态和瞬态过程。采用虚拟场景技术,按高温气冷堆的实际结构建立三维虚拟场景,用户可在虚拟场景中漫游观测,实时查看仿真计算状态;同时可对仿真数据结果进行分析并以二维、三维图形显示。该仿真系统不仅对高温气冷堆的工程设计、安全分析和人员培训有重要作用,且可以对HTR-10主控室的操作人员进行现场支持及各项研究提供帮助。     

高温气冷堆PD输出反馈功率控制的镇定性能分析

由于具有固有安全性,高温气冷堆被视为下一代核能系统的首选堆型之一,且安全、稳定和高效的运行是发展高温气冷堆技术的基本要求。功率控制通过合理给出控制棒的升降速度来增强闭环稳定性和过渡过程暂态性能,因而对于提升高温气冷堆的运行性能具有重要意义。本文基于反应堆控制的物理方法,从理论上给出了比例微分(PD)输出反馈功率控制律保证闭环全局渐近稳定的充分条件。数值仿真结果验证了此结论的正确性,并揭示了功率调节性能与温度反馈回路增益之间的关系。这一结果不仅从理论上保证了经典PD反馈律可用于实现高温气冷堆的高性能功率调节,而且为利用PD控制实现负荷跟踪提供了理论基础。     

10MW高温气冷堆蒸汽安全阀全性能试验

文章介绍10MW高温气冷堆(HTR-10)二回路超压保护系统中的核二级蒸汽安全阀的设计要求、结构特点及性能要求,并对其性能进行了实验验证。实验结果表明：蒸汽安全阀的性能满足设计要求,达到了核规范的标准。     

10MW高温气冷堆蒸汽发生器双管工程模拟实验装置

介绍了１００ＭＷ高温气冷堆（ＨＴＲ－１０）蒸发发生器双管工程模拟实验装置实验回路及主要实验设备的技术特征和主要技术指标，该实验装置用两根螺旋蒸发管作为实验本体，用高温氦气作为热源，全部采用全尺寸模拟。实验回路由氦气回路，一次水回路，二次水回路组成。一次侧氦气的工作压力为３．０ＭＰａ，工作温度为６７０℃二次测蒸汽压力为４．０ＭＰａ，工作温度为４４０℃。该装置主要研究ＨＴＲ－１０蒸汽发生器３０％负荷运     

HTR—10石墨球与燃料球均匀混合装料初装堆方案研究

分析了球床式高温气冷堆几种可能的初装堆方案的特点，选取石墨球与燃料球均匀混合作为１０ＭＷ高温气冷实验堆的初装堆方案。利用高温气冷堆物理设计程序ＶＳＯＰ进行计算，分析屯ＨＴＲ－１０从初始装料向平衡态过渡过程中的倒换料方式，最大单球功率及最大燃耗变化情况。     

HTR-PM蒸汽发生器二次侧吹扫技术研究

基于国内高温气冷堆示范工程(HTR-PM)蒸汽发生器(SG)二次侧的清洁问题,提出压缩空气吹扫技术。通过对SG二次侧螺旋管汽-液两相和单向流特性的分析,得到吹管系数的计算方法。根据HTR-PM机组的设计参数进行实例计算,得出压缩空气吹扫可行性方案。     

高温气冷堆闭式布雷顿间接循环中氚的来源及其影响

氚是氢的放射性同位素,影响环境和人体健康.目前,全球自然界中的氚主要来自人类的核活动.因此,需研究核反应堆中氚的来源及其影响.在高温气冷堆中,氚是一回路放射性的主要来源之一.由于高温气冷堆堆芯温度较高,不能忽视一回路中氚向外界和二回路渗透造成的污染问题.文章阐述了氚的物理和化学特性,高温气冷堆闭式布雷顿间接循环中氚的生成来源和释放途径,分析了氚对设备材料力学性能的影响,介绍了氚向环境释放的限值、控制措施及防止氚渗透的方法.     

A decade''s experience in chemistry maintenance of once through steam generator of fast breeder test reactor

The steam generator of fast breeder test reactor (FBTR) at Kalpakkam (India) is a once through Steam Generator (OTSG) which requires the feed water at high purity level. Therefore, for maintaining feed water chemistry, all volatile treatment (AVT) is adopted along with a full-flow deep bed Condensate Polishing Unit (CPU) in the steam water system. Operational difficulties such as premature termination of operation cycle of the CPU, enhanced impurity pickup resulting in increased load for CPU, early silica breakthrough, etc. were observed on occasions. This paper describes the modifications carried out in the steam water circuit to overcome these problems. A decade's experience in operating the CPU and maintaining the feed water quality is also discussed.     

Transport of nuclear heat by means of chemical energy (nuclear long-distance energy)

Nuclear long-distance energy, i.e. the transportation of chemically bound energy, represents a potential application for process heat plants in which the endothermic reaction takes place at the heat source (high temperature reactor) whereas the exothermic back reaction occurs at the region of heat utilization (consumer). Due to the following criteria, i.e. reversibility of the chemical reaction, sufficiently large reaction enthalpy, favourable temperature region for the forward and back reactions, and the available technology, a combination of the methods of endothermic steam reforming of methane and exothermic methanation is chosen. As well as supplying household and industrial consumers with heating, process steam and electrical energy, an interconnected system with synthesis gas consumers (e.g. methanol production and iron ore reduction plants) is possible. It is shown that the amount of reactor heat which is convertible into long-distance energy depends considerably on the helium temperatures in the high temperature reactor and lies between 60 and 73% of the reactor power. Conceivable circuit schemes for the nuclear steam-reforming plants and the methanation plants are described. Finally, it is demonstrated, with the help of a simple model for cost estimations, that the nuclear long-distance energy system can make heating for households available in competition with oil heating and that due to the lower specific transport costs, for distances larger than 50 km it is also more economical than the hot water supply from the thermal power coupling of steam turbine plants using light water reactors (LWRs) or high temperature reactors (HTRs).     

示范快堆给水控制系统仿真研究

针对示范快堆每个环路设置多个直流式蒸汽发生器模块的特点，提出集中控制环路给水流量和单独控制环路内各模块给水流量的控制方案，并搭建了给水控制系统的仿真模型；分别对2种给水控制方案进行仿真研究，分析在2种给水控制方案下各模块蒸发器出口钠温和蒸汽过热度的变化规律。研究结果表明:集中控制环路给水流量的控制方案更有利于保证蒸发器出口蒸汽过热度的安全限值，而单独控制环路内各模块给水流量的控制方案更有利于蒸发器出口钠温的控制。      

自主化1000MW级压水堆核电站核蒸汽供应系统概念设计

本文简介介绍了我国百万千瓦级压水堆核电站（ＣＮＰ１０００）核蒸汽供应系统的概念设计，主要内容为主要技术参数、堆芯设计、反应堆冷却剂主回路系统及其主要设备设计、安注系统、辅助给水系统和数字化仪表与控制系统设计。     

Probabilistic assessment of the fracture mechanics behaviour of an HTR-module primary circuit pressure boundary

In this paper the fracture mechanical behaviour of the primary circuit pressure boundary of a planned HTR-module reactor for electricity and steam generation under normal operation is assessed probabilistically. First and second order reliability methods (FORM-SORM) are used to calculate failure probabilities. They also allow a simplified analysis of the leak-before-break (LBB) behaviour. No LBB was probabilistically identified for the reactor pressure vessel. However, failure of the pressure vessel unit in normal operation probably originates from the connecting pressure vessel or the steam generator pressure vessel. They show LBB in probabilistic terms.     

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.     

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.     

Design and evaluation of Operation Grade Decay Heat Removal System of PFBR

The decay heat removal (DHR) system removes the decay heat generated (by radioactive decay of fission products) in the core after the reactor is shut down, thereby ensuring proper cooling of the core sub assemblies and limiting main vessel, internals and sodium temperature within safe limits. There are two diverse paths for removal of decay heat from the reactor, namely, Safety Grade Decay Heat Removal System (SGDHRS) and Operation Grade Decay Heat Removal System (OGDHRS). OGDHR circuit is used when at least one secondary sodium loop, DHR related steam water circuit and off site power supply is available and SGDHR circuit is used when OGDHR system is not available or when both the secondary loops are not available for DHR. This paper provides brief details of the design and evaluation of OGDHRS.     

超临界水在垂直管内换热及流动不稳定性研究

清华大学核能与新能源技术研究院在建的250 MWt高温气冷堆核电站示范工程(HTR-PM)中蒸汽发生器二回路为亚临界水,由于反应堆能提供750℃的高温氦气,二回路水可提高到超临界压力和温度,采用多堆带一机方案可与超临界蒸汽透平机组匹配,因此研究超临界水在管内的流动、传热以及流动不稳定现象非常重要。本文通过使用RNGk-ε模型耦合强化壁面函数,发现模拟结果与Yamagata等的实验数据符合较好。基于此模型,分析了超临界流体流动时换热系数的变化规律,并采用瞬态计算方法,线性增大加热功率,分析了流动不稳定现象,发现流体一旦进入不稳定区,进出口流量的波动非常严重,甚至出现倒流,应尽可能避免此类现象。