水力控制棒驱动系统失压工况的实验研究

利用200MW低温核供热堆水力控制棒驱动系统的1：1实验台架模拟系统失压况，进行控制棒步升，步降，开阀落棒及关泵开阀落棒实验，并与正常工况下的提棒，落棒的实验结果进行比较，实验结果表明：在系统失压工况下，控制棒能正常提棒，落棒系统失压工况下的弹棒，系统压力与压力壳压力具有瞬时跟随特性，未出现控制棒弹棒事故，对实验的失压速率和事故分析得到的失压速率进行了比较，验证了系统具有良好的安全性和可靠性。     

间接氦气透平循环高温堆紧急停堆动态仿真

通过对10 MW高温气冷堆氦气透平发电装置(HTR-10GT)的堆芯、热交换器和透平压气机组等主要设备的数学建模和程序编制,初步建立起了一套模拟该装置瞬态特性的仿真程序.通过对该装置于5s时刻堆内引入0.1$阶跃正反应性引发的紧急停堆事故的瞬态模拟,初步验证了该装置紧急停堆预案设置的安全性和合理性,证明了旁路快开阀的设...     

49-2游泳池式反应堆超设计基准事故的筛选与分析

为保证49-2游泳池式反应堆在超寿期下的安全运行,需进行超设计基准事故分析。由于难以采用概率安全评价(PSA)方法进行分析,所以本文无条件假设最严重事故来得到一保守结果。主要分析了全厂断电下未能紧急停堆的预期瞬变(ATWS)、水平孔道断裂和停堆后堆芯完全裸露的事故,以及应急能力。结果表明:在全厂断电ATWS下堆芯是安全的;水平孔道断裂及其他因素造成失水时,只要2.5h内堆芯不裸露即可保证燃料元件不熔化;非能动破坏虹吸能力和多样的应急补水方式能保证堆芯不裸露。     

机组临时停机停堆期间的化学监督与控制

目前的化学与放射化学程序和措施,都是针对正常功率运行和按部就班有计划的大修状态而设置,如遇到机组跳堆、跳机或冷停堆等紧急情况,则没有相应的应急预案或相关程序进行提前或有目的地干预。基于这种情况,电厂化学人员经过多年的实践和不断经验反馈,总结并编写了专门针对紧急停机停堆的化学监督与控制应急预案。通过停堆过程和停堆后的不同状态,启机过程的化学与放射化学监测,监督燃料包壳状态,控制一回路的剂量水平,以防止设备腐蚀。     

空间核动力装置控制鼓系统试验样机热态性能试验

控制鼓系统是空间核动力装置上执行功率调节、紧急停堆的重要安全设备,其能否正常运行直接关系到核动力装置的安全性。为验证控制鼓系统能否满足设计要求,必须进行热态下的性能试验。本文采用1∶1全尺寸控制鼓系统试验样机,通过设计建立专用的热态性能试验装置,对试验样机寿期内全行程往复、电机切换和快速复位功能进行了试验验证和研究分析。试验过程显示,试验样机运行基本平稳,无异响和卡顿,快速复位时间满足设计指标,但传动链终端存在角度滞后、旋转过程位置重复精度低和小角度快速复位乏力等现象。该控制鼓系统试验样机机构设计基本满足机械运转功能,为下一阶段控制鼓系统结构的优化与定型奠定了基础。     

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。     

Safety Analysis of Abnormal Reactivity Events in HTTR

Abstract

To confirm the safety of the High Temperature Engineering Test Reactor (HTTR) facility which is being constructed as the first high temperature gas cooled reactor in Japan, the representative abnormal reactivity events assumed in the safety analysis of the HTTR were analyzed. The HTTR is a graphite moderated and He-gas-cooled reactor with thermal power of 30 MW, inlet coolant temperature of 395°C and outlet coolant temperature of 950°C.

This report presents the analytical results of two representative events, “Abnormal control rod withdrawal from a subcritical condition” and “Abnormal control rod withdrawal during the full power operation”, showing that the safety of the HTTR is secured in conformity with the unique features of the HTTR with respect to the maximum fuel temperature, which is a key factor for the safety criteria.

The results of the safety analysis could demonstrate the safety of the HTTR facility with respect to abnormal reactivity events postulated in the HTTR, showing that the maximum fuel temperature is lower than the limit of the maximum fuel temperature of 1,600°C.     

A scenario of core disruptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency (JAEA) has selected the advanced loop-type fast reactor, so-called JSFR. The safety design requirements of JSFR for Design Extension Condition (DEC) are the prevention of severe accidents and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, in particular, the In-Vessel Retention (IVR) against postulated Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the in-vessel cooling of degraded core materials are evaluated so as to achieve IVR. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulations. This is an unprecedented approach to the construction of a CDA scenario, and is an effective method to objectively investigate the factors leading to IVR failure and the design measures against them. It was concluded that mechanical/thermal failures of the reactor vessel due to power-excursion/thermal-load could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.     

基于SAC-3D系统软件的FFTF未能紧急停堆失流实验的数值模拟

本文利用系统分析软件SAC-3D对美国快通量试验堆（FFTF）堆芯及一回路进行了建模,并根据国际原子能机构（IAEA）提供的FFTF未能紧急停堆的失流实验的边界条件数据进行了事故瞬态仿真计算。计算得到堆芯热工水力及中子物理关键参数,仿真结果与实验测量数据符合较好。对比结果验证了SAC 3D在模拟液态金属冷却快堆事故工况中的有效性与准确性,也证明了FFTF堆型具有可靠的非能动安全性。     

中国实验快堆事故停堆后余热排放过程的数值模拟

分析了中国实验快堆事故停堆后余热的排放过程。对热钠池中的流动与传热采用多孔介质模型的全三维数值模拟,对堆芯支路、事故热交换冷却回路和空冷塔冷却支路采用一维系统分析程序进行数值模拟。通过三维部分和一维部分相互耦合,模拟了余热排放的瞬态过程,得到了堆芯出口温度、燃料元件包壳的最高温度、余热热交换器的余热排放功率等许多重要参数随时间的变化曲线,对中国实验快堆的安全设计有重要的参考价值