主泵参数变化对压力容器液位测量影响分析

为了量化分析CPR1000核电厂主泵特性及相关参数（如电网频率、空泡份额等）变化对堆芯冷却监测系统（CCMS）压力容器液位(L VSL)测量引入的误差,评价该误差对事故处理进程的影响,基于CCMS L VSL测量原理,推导出主泵各参数变化对L VSL测量引入误差的计算公式,并进行量化计算。计算结果表明,除主泵本身的性能降级会导致L VSL较大的低估误差外,其余参数变化对L VSL测量引入的误差可忽略。结合状态导向法事故运行程序（SOP）,分析了主泵本身性能降级导致的低估误差对操纵员关键安全操作的影响。结果表明,该误差可能干扰SOP中主泵的相关操作,但不会阻碍SOP事故处理中关键安全操作的执行。     

堆芯温度分布对压力容器液位测量影响分析

对堆芯温度不均匀分布而导致CPR1000核电站堆芯冷却监测系统CCMS压力容器液位测量误差进行了量化计算。结果表明,停堆后主泵保持运行,由该物理现象引入的误差可以忽略。对失去全部给水情形下引入较大的高估误差,结合状态导向法事故运行程序SOP,对该误差对操纵员安全重要操作的影响进行了分析。     

CPR1000核电站堆芯出口冷却剂过冷度测量不确定度评定

中国百万千瓦级先进压水堆(CPR1000)核电站反应堆通过堆芯冷却监测系统(CCMS)测量堆芯出口冷却剂的过冷度。本文分析了堆芯出口冷却剂过冷度测量过程中的各种误差来源,对饱和状态下堆芯出口冷却剂温度测量的不确定度进行评定,得到不确定度区间边界随一回路压力变化的曲线,给出了用于判断堆芯冷却状态的堆芯出口冷却剂过冷度测量的误差ε曲线的确定方法,该方法已在CPR1000核电站中得到实际应用。     

LOCA后物理现象对压力容器水位测量影响分析

为了验证CPR1000核电站冷却剂失流事故（LOCA）下堆芯冷却监测系统（CCMS）压力容器水位（L VSL）测量的有效性,对LOCA后影响动压和静压测量的物理现象,以及顶盖的特殊现象对L VSL测量引入的误差进行了量化计算。结果表明,顶盖的特殊现象和冷管段或热管段破口流量对L VSL测量引入的误差在破口发生几分钟后可忽略,压力容器顶部破口流量以及控制棒导向管内水的滞留对L VSL测量引入很大的高估误差,结合状态导向事故处理程序SOP的分析表明,该高估误差不会阻碍事故处理安全重要操作的执行。     

CPR1000核电机组堆芯水位监测系统及其调试

CPR1000核电机组反应堆堆芯水位监测系统是反应堆发生LOCA事故后监测堆芯淹没状态的重要系统,由其测量的水位直接用于反应堆事故规程的导向。本文对该系统的测量原理、系统构成进行了详细的介绍,通过对CPR1000核电机组首台机组的调试,实现了该系统的首次自主化调试的目标。     

堆芯流量变化对压力容器水位测量影响分析

CPR1000核电厂在每次换料大修期间需执行CCMS(Core Cooling and Monitoring System)校验试验,以获得计算压力容器水位L_(VSL)所需的堆芯动态压头损失系数,完成该试验耗时较长。论文依据调试和换料大修期间一回路冷却剂流量的变化情况评估堆芯动态压头损失系数的变化,并定量评价对L_(VSL)测量的影响。分析结果表明,在回路水力特性未发生明显变化的情形下,对L_(VSL)测量引入的误差很小。建议在L_(VSL)测量不确定度评定时引入堆芯动态压头损失变化的影响,在换料大修时校验流量变化对堆芯动态压头损失的影响是否在允许范围之内,可简化CCMS校验试验,提升机组的经济性。     

AP1000核电厂丧失主给水ATWS事故CMT注射分析

堆芯补水箱（CMT）是AP1000核电厂非能动堆芯冷却系统（PXS）的重要组成部分。在通常情况下,当主泵开启时,CMT即使被触发,也不能注入堆芯。然而在某些事故工况下,即使主泵开启,CMT也有可能注入,它将直接影响事故进程及分析结果。应用压水堆核电厂通用系统程序RELAP5MOD3.1对AP1000核电厂丧失主给水ATWS事故进行了计算分析,验证了美国西屋公司LOFT4AP2.0.1程序计算结果的正确性,并分析找出了CMT成功注入的根本原因。     

大亚湾核电站堆芯冷却监测系统改造

针对大亚湾核电站堆芯冷却监测系统(CCMS)面临部件老化、备件无法采购导致系统工作不稳定及故障报警闪发的现状,提出对CCMS进行整体升级改造。描述新CCMS采用国产化安全级仪控平台Firmsys的设计方案,以及对新CCMS的功能及接口等关键技术进行研究与分析;介绍在安装调试阶段遇到的技术问题及其解决方法。对改造后系统进行功能验证,以实现CCMS的自主化设计和改造。     

全厂断电事故下封头熔池传热行为的研究

利用SCDAP/RELAP5系统程序对CPR1000核电厂进行了建模,并对全厂断电事故(SBO)的进程进行了模拟,分析了SBO中从堆芯开始裸露到完全裸露的熔化过程以及堆芯熔融物掉入下封头后下封头中熔池的传热行为。结果表明,熔融物在下封头形成一个混合层和重金属多孔介质层,且失效的位置在下封头侧部30°～40°位置(压力容器底部为0°)。     

“华龙一号”堆芯冷却监测系统设计

“华龙一号”是我国自主研发的第三代核电站,其反应堆及一回路系统在设计中对固有安全性提出了更高的要求。对于二代加核电厂堆芯冷却监测系统（CCMS）,需要在反应堆底部开孔测量水位。该设计降低了反应堆固有安全性,必须重新设计。本文设计了一种新型CCMS,其探测器从压力容器顶盖插入堆芯进行直接测量,不但提高了关键点的水位测量准确度,同时避免了压力容器底部开孔,满足了“华龙一号”反应堆固有安全性要求。      

Coolant boiling noise in LMFBRs

The noise that may arise due to boiling of Na in LMFBRs is investigated. The study deals specifically with investigations of the likely frequency range of Na boiling and its dependence on local parameters like coolant flow velocity, cavity size and contact angle.     

主泵参数变化对失水事故后果影响分析

基于100D主泵和ANDRITZ主泵的差异,分析主泵相似特性曲线和自由容积的变化对失水事故(LOCA)后果的影响。针对岭澳核电站二期反应堆冷却剂系统,应用CATHARE GB程序和CONPATE4程序分析大破口LOCA事故堆芯热工水力后果;应用ATHIS和FORCET程序分析失水事故喷放阶段的反应堆冷却剂主管道水力载荷。结果表明,主泵相似特性曲线的变化对大LOCA事故再淹没阶段的堆芯热工特性影响很大,采用不同主泵时的最高峰值包壳温度(PCT)相差很大;而主泵自由容积对失水事故喷放阶段的卸压波传递影响较大,导致采用不同主泵时的反应堆冷却剂主管道水力载荷有所不同。     

反应堆主冷却泵的冷态调试

文中介绍了秦山核电厂反应堆主冷却泵冷态调试的内容及试验方法。主冷却泵的现场冷态试验包括:报警试验、联锁试验和运转试验。各项试验的结果表明,主泵的运行参数正常,联锁、报警和性能符合设计要求。     

考虑管路中冷却剂动能的反应堆冷却剂泵惰转计算模型

核主泵惰转惯量设计过小,一旦核电站全厂停电会造成核事故,而设计过大会极大地降低机组效率,因此惰转计算模型的准确性对于保证核电站安全和提高机组效率十分重要。本文考虑管路中冷却剂动能对反应堆冷却剂泵惰转过程的影响,通过启-停机过程中功率守恒方程和泵相似定律,推导并建立了考虑管路冷却剂影响的惰转瞬态计算模型,并给出了泵机组惰转惯量和惰转时间的简单计算公式,使计算结果更精确,工程适用范围更广泛,可应用于核工程和非核工程中惰转惯量的精准设计以及惰转时间的精准计算。      

Coolant stratification in nuclear power facilities

Computational and experimental studies of the structure of the motion of stratified coolant and its temperature regime are performed on models of different elements of the circulation loops of fast and thermal reactors. The investigations have shown that in the presence of stable stratification thermogravitational forces bring about the formation of stagnant and recirculation formations with large temperature gradients and pulsations at interfaces. The data obtained show that stratification phenomena must be taken into account when validating the reliability of the control, safety, and nominal service lives of nuclear power facilities.     

Danger of Burning of Sodium Coolant

The properties of sodium as a substance that can burn are described. The radiation characteristics of sodium as a first-loop coolant in a fast nuclear reactor are presented. An assessment is made of the consequences of sodium burning in various situations. First and foremost, an unanticipated accident with burning of sodium in the first loop adopted in the BN-800 design, is examined. Next, situations with hypothetical scenarios are examined to obtain the limiting data charcterizing the potential fire hazard of radioactive sodium coolant. Specifically, a hypothetical situation where all of the sodium contained in the first loop of the reactor burns is examined. The computational results are analyzed from the standpoint of the role sodium plays in the overall problem of nuclear power plant safety.     

A Dewpoint Device for HTGR Coolant

High-temperature gas-cooled reactors (HTGRs) use a gaseous coolant for heat transfer between the nuclear core and two or more steam generators. Leakage of steam or water from the steam generators to the coolant would expose the nuclear core to water vapor. A moisture measuring system is required to determine the moisture content of the coolant gas in the range of 0.1 to 3000 volume parts per million (ppm). Another requirement is the rapid detection of large leaks resulting in 2000 ppm or more and the identification of the leaking steam generator, thus permitting isolation of the faulty coolant loop. An optical dewpoint detector has been developed that can be used either as a dewpoint monitor or as a dewpoint trip device. The response time of the device as a trip instrument is typically 1 sec in the dewpoint range of 27°F to 128°F (100 to 3000 ppm). As a dewpoint monitor, the mirror temperature can be changed at a rate of 1°F/sec, in the range of -87°F to +128°F (0.1 to 3000 ppm). The moisture detector head is designed to operate at the full coolant pressure of 700 psia. In the HTGR application, access to the device is difficult during reactor operation, and will be cumbersome at all times because of gamma radiation environment. Therefore, exhaustive testing of all detector head components, subassemblies, and materials selection from inorganic substances has been performed to reduce maintenance to a minimum.     

Coolant cavitation in dynamic containment loading

After a brief introduction to the subject of cavitation in subcooled liquids and a survey of what is known regarding the key parameters in the cavitation process for water and for sodium, the basic equations of the SIMON cavitation model for use with Lagrangian containment codes and the assumptions behind them are reviewed.Some calculations using this model are then presented which show the dissipative effect of cavitation both in uncavitated liquids transmitting tension waves and in cavitated liquids transmitting pressure waves. The cavitation which develops when pressure waves are reflected at free surfaces is also examined, and some calculated results are compared with an experiment involving this phenomenon found in the literature. The role of cavitation in the containment loading process is then discussed, and examples taken from model test calculations are adduced to show that cavitation occurs at all stages of the loading process and involves a high proportion of the total liquid volume. Again by example the point is made that in certain simple circumstances a crude pressure cut-off model of cavitation is adequate but that for other major aspects of the containment loading process such as roof impact pressures and structural deformations a more refined model is necessary.