高温气冷堆热气导管的结构稳定性分析

热气导管是高温气冷堆中氦气循环的重要流道,其在各工况下的结构完整性与稳定性关系到反应堆是否能运行安全。本文详细分析了热气导管在事故工况下所承受的压力载荷,包括绝热纤维对管壁的压力以及一回路失压事故时发生氦气泄漏产生压力;并根据得到的压力载荷计算了热气导管承受外压时的结构稳定性。计算结果表明热气导管在事故工况的压力载荷作用下能够保持结构的完整性并且不会发生失稳。     

HTR-PM蒸汽发生器绝热层支撑结构的设计与分析

为提高HTR-PM高温气冷堆蒸汽发生器的热交换效率并分隔冷热氦气的通路,必须在蒸汽发生器中设置绝热装置.在围绕换热管放置的钢制围筒上铺设一定厚度的绝热纤维,用多块覆盖板将绝热纤维压紧,在每块覆盖板中心与四角,用中间填充绝热纤维的螺栓将覆盖板定位并紧固,组成绝热层支撑结构.使用有限元方法对绝热层支撑结构在绝热材料压力、蒸汽发生器失压、地震等多种载荷的综合作用下的应力分布情况进行计算,计算结果表明:支撑结构在多种载荷的作用下仍能保持绝热层功能的可靠性与结构的完整性.     

蒸汽发生器地震反应分析及参数敏感性研究

蒸汽发生器是排出反应堆堆芯产生热量的主要设备,是反应堆冷却剂系统压力边界的一部分。其为抗震I类设备,须对其进行地震反应分析。本文建立了蒸汽发生器地震反应分析模型,地震反应分析模型包含汽水分离器组件和管束组件等内件。两个蒸汽发生器模型与一回路管道和压力容器串联,进行地震反应分析,获得地震载荷下的应力结果。同时,本文还就地震反应分析结果对各参数的敏感性做了研究,其中包括另一台蒸汽发生器、支撑、抗振条设置等的影响。研究结果表明,地震反应结果对设备支撑和抗振条设置特别敏感。本文总结了这些参数对分析和设计的指导性意见,供后续核电站蒸汽发生器设计和研发时参考和关注。     

10兆瓦模块式高温气冷实验堆的设计概念及安全分析

概述了将在我国建造的10兆瓦模块式高温气冷实验堆的设计特性、安全概念及安全分析。介绍与讨论了部分设计基准事故的分析结果,包括反应性引入事故、丧失主热阱事故、主回路失压事故,以及蒸汽发生器断管事故等。初步安全分析结果表明,10兆瓦高温气冷实验堆具有良好固有安全特性,在上述假想事故工况下反应堆均处于安全允许状态。     

10MW高温气冷堆热气导管高温性能试验

水平布置的同轴双层热气导管在１０ＭＷ在高温气冷实验反应堆中是连接堆芯和蒸汽发生器的重要部件， 外分别流过高温和低温氦气，在氦气工程试验回路上进行了热气导管热工性能试验，使用氦气介质，在３．０ＭＰａ，９５０℃温度下连续运行时间超过９８ｈ，ｄ３．０ＭＰａ７００℃以上温度条件下的热运行时间超过３５０ｈ，还在０．１－３．４ＭＰａ压力范围内进行了２０闪压力循环；在１００－９５０℃范围内，进行了１８次温度循     

反应堆冷却剂系统动力分析关键参数的敏感性分析

反应堆冷却剂系统(RCS)在极端事故工况下的动力响应计算是评估核电厂安全的重要技术手段。定量考察系统结构的关键参数对系统动力响应的敏感性,是可靠评价系统响应的重要方面。本文通过全局敏感性和相关性分析,对一种堆型蒸汽发生器(SG)支承刚度对地震条件下主系统载荷分配的敏感性进行了研究。研究表明,支承刚度对SG局部范围内主系统载荷分配影响度较高,对距离较远的反应堆压力容器影响度较低。此外,还建立了描述关键参数到载荷分配的输入输出关系,并通过神经网络对输入输出关系进行了回归建模。该神经网络模型能够快速准确地对发生支承结构设计变更后的主系统地震载荷分配进行评估。     

基于PIPENET的AP1000主蒸汽管道汽锤计算

本文通过汽锤力载荷的简化计算方法求解汽锤载荷力。应用PIPENET软件对国内某三代核电厂蒸汽发生器到主蒸汽母管之间的蒸汽发生器系统进行建模。利用瞬态计算功能模拟汽锤发生及衰减过程,给出最大汽锤压力、管系中最大汽锤载荷、该载荷发生的时间及管道位置。并与西屋公司的管道应力计算文件进行对比分析,验证了该计算方法。     

压水堆蒸汽发生器传热管完整性要求

１ 相关标准１．１　美国联邦法规１０ＣＦＲ５０要求符合５０．２１（ｂ）或 ５０．２２的运行许可证持有者必须制定和执行一个保证大纲,该大纲要保证压水堆蒸汽发生器传热管的安全功能。首要的安全功能是由于蒸汽发生器传热管是反应堆冷却剂压力边界（ＲＣＰＢ）的主要组成部分,必须要保持反应堆冷却剂的总量和压力。其次,蒸汽发生器传热管作为一、二回路之间的热交换导热体,还保证了反应堆的停堆能力。第三,蒸汽发生器传热管隔离了一回路系统里的放射性介质,避免它们进入二回路系统和释放到环境中去。 制定传热管完整性大纲是为了…     

核反应堆地震和失水事故下结构动力分析研究

对核反应堆在地震和失水事故下结构动力分析开展了研究,包括反应堆和蒸汽发生器在失水事故下的瞬态卸压载荷分析研究、反应堆结构动力响应分析研究和燃料组件动力响应分析评定研究.形成了一套可用于新核电站反应堆结构设计的完整的设计分析和计算软件系统,并已在秦山核电二期工程和CNP1000反应堆结构设计中得以应用     

蒸汽发生器堵管对反应堆安全的影响

在核电站中,蒸汽发生器的部份U形管常常需要堵管。本文分析了蒸汽发生器的部份U形管被堵后继续运行对反应堆热工设计安全性能的影响,并予计这种影响的程度;文内还讨论了与此相关的一些问题;最后简略地提出了对蒸汽发生器的设计改进意见。     

Design and experiment of hot gas duct for the HTR-10

A horizontal coaxial double-tube hot gas duct is a key component connecting the reactor pressure vessel and the steam generator pressure vessel for the 10 MW High Temperature Gas-cooled Reactor—Test Module. Hot helium gas from the core outlet flows into the steam generator through the liner tube, while helium gas after being cooled returns to the core through a passage formed between the inner tube and the duct pressure vessel. Thermal insulation material is packed into the space between the liner tube and the inner tube to resist heat transfer from the hot helium to the cold helium. The thermal compensation structure is designed in order to avoid large thermal stress because of different thermal expansions of the duct parts under various conditions. According to the design principal of the hot gas duct, the detailed structure design and strength evaluation for it has been done. A full-scale duct test section was then made according to the design parameters, and its thermal performance experiment was carried out in a helium test loop. With helium gas at pressure of about 3.0 MPa and a temperature over 900 °C, the continuous operation time for the duct test section lasted 98 h. At a helium gas temperature over 700 °C, the cumulative operation time for the duct test section reached 350 h. The duct test section also experienced 20 pressure cycles in the pressure range of 0.1–3.4 MPa, 18 temperature cycles in the temperature range of 100–950 °C. Thermal test results show an effective thermal conductivity of the hot gas duct thermal insulation is 0.47 W m⁻¹ °C⁻¹ under normal operation condition. In addition, a hot gas duct depressurization test was carried out; the test result showed that the pressure variation occurred on the liner tube was not more than 0.2 MPa for an assumed maximum gas release rate.     

HTR-10一体化换热装置和联合循环

10MW高温气冷实验堆 (HTR 10 )有两个突出的特点 :(1)采用气体轮机和汽轮机联合循环 ,可以提高其发电效率 ,并且能保持较低的堆芯入口温度 (30 0℃ )。 (2 )中间换热器和蒸汽发生器一体化设计 ,减少了一个壳体和一个热气导管 ,提高了安全性和经济性。描述了一体化的中间换热器和蒸汽发生器以及气体轮机和蒸汽轮机联合循环的设计特点。     

The primary loop confinement and pressure boundary system of the HTR-10

The design philosophy and requirements of the HTR-10 reactor building and the primary loop confinement are introduced in this paper. Also introduced are the design, fabrication and the installation of the HTR-10 primary loop pressure boundary system. The primary loop confinement comprises the sealed cavities of the reinforced concrete structure. The main components and the connected gas systems of the primary loop pressure boundary system are contained in the confinement. Under normal operating condition, the inside pressure of the confinement is kept at negative pressure to ensure the sealing function of the confinement. There is a rupture disk of overpressure protection in the confinement wall. After a depressurization accident the pressure of the confinement increases and the rupture disk will break. The air of the confinement is discharged directly to the atmosphere through the accident discharge chimney which is connected to the rupture disk without filter. The main components of the primary loop pressure boundary system consist of the reactor pressure vessel, the steam generator pressure vessel and the hot gas duct vessel. All the above main components are installed in the reactor cavity and the steam generator cavity. They are all nuclear safety class 1 components, whose materials production, design, fabrication, and tests are carried out according to ASME Section III and relevant Chinese nuclear codes.     

低参数直流蒸汽发生器稳定性实验研究

针对核供热堆的低温、低压运行条件，建立了大盘管直流蒸汽发生器热工水力学实验系统，实验研究了低参数条件下入口流动阻力、运行负荷和蒸汽发生器出口压力对直流蒸汽发生器稳定性的影响。结果表明：入口节流阻力比、运行负荷、系统压力对直流蒸汽发生器稳定性有影响，通过调整入口节流阻力比，直流蒸汽发生器可在低压参数下稳定运行。     

高温气冷堆氦气透平直接循环启动参数分析计算

开发了包括堆芯、蒸汽发生器、透平、压气机及换热器等模块在内的高温气冷堆氦气透平直接循环系统的稳态计算程序。对系统的启动过程进行了模拟分析,并对压气机的喘振问题进行了分析,考虑了换热能力、温度和压力的影响。结果表明：在变负荷过程中压气机有足够的安全裕度。     

高温堆热气联箱内部流场分析

以高温堆热气联箱为研究对象,在实验研究的基础上,采用流体力学计算程序CFX5对热气联箱和热气导管内部流场进行了数值模拟,以获得热气联箱和热气导管内的速度场、压力场和温度场,为高温堆热气联箱的设计和实验研究提供参考.数值计算结果表明:热气联箱内气流发生剧烈搅混,加速了不同温度气流间的热传递,有利于高温和低温气流间的温度混合.但存在肋片的区域没有发生剧烈的气流搅混,不利于气流间的热传递.热气导管内温度混合率随其长度的增加逐渐增大,热气导管长度2.5m以上时,温度混合率达到99%以上.     

Numerical study on thermo-hydrodynamics in the reactor internals of SMART

The present study investigates distributions of pressure and temperature in the SMART, which is one of the nuclear reactors developed by KAERI. The pressure, the velocity, and the temperature are calculated under nominal operation and non-uniform temperature distributions are predicted in the inlet of core generated by the cases of broken-down steam generator. 3D-numerical simulations using a FVM commercial code is performed to calculate the distributions. Under nominal operation, the pressure drop is the highest in the steam generators as well as the temperature decrement is the highest. Broken-down steam generator yields the unmixed flow. When the steam generator related to the fourth floor in the flow mixing header assembly is broken down, the unmixed flow generates the narrowest hot flow region and the widest temperature range at the inlet of the reactor core.     

Air—water model studies of cocurrent flow into and along a PWR hot leg to the steam generator

An experimental study is made of the cocurrent flow of air and water at atmospheric pressure from a reactor vessel into and along an approximately ninth-scale replica of the Sizewell ‘B’ PWR hot leg to the steam generator. A flow regime map of conditions in the hot leg is presented.The water interface level in the reactor vessel as a function of the flowrates is in agreement with a recent theory developed by the author. The same theory predicts the level in the hot leg when discharging two phases through a horizontal break and is in agreement with the results of other workers on this subject for the discharge of air and water up to a pressure of 5 bar and of steam and water up to a pressure of 62 bar.Results on the water level in the hot leg are correlated empirically but, for lower flowrates, the results are in approximate agreement with a theory for the onset of flooding.     

Component exposure in hypothetical accidents with very fast depressurization in a HTR module reactor

The starting event of the massive air ingress into the core of the HTR module reactor, classified as hypothetical incident, is the very fast depressurization of the primary circuit. Provided that the integrity of the reactor pressure vessel is not in question, a rupture of the connecting pressure vessel between reactor pressure vessel and steam generator vessel is the maximum possible leak cross-section. In this work it is investigated whether the components of the reactor pressure vessel are exposed by the depressurization process to mechanical loads which exceed the load limits. These loads are caused by two different events, the strong momentum change of the fluid and the local pressure differences, respectively. Due to the momentum change the bottom reflector receives the maximum load, whereby only 2% of the compressive strength of the graphite quality used there are reached. However, the load by local pressure differences is between passed volumes and in normal operation, not-passed volumes lead to high load values. A maximum pressure difference of 44.5 bar was calculated at the thermal top shield.     

船用压水堆核动力装置双恒定运行方案静态特性研究

讨论了船用压水堆核动力装置的双恒定运行方案以及实现的技术手段 ,并通过反应堆热工安全准则的计算和蒸汽发生器传热实验 ,从稳态运行过程的角度探讨了船用核动力装置实现双恒定运行方案的可行性。