蒸汽冷凝回流模型

蒸汽冷凝回流冷却是压水反应堆发生失水事故时的一个重要堆芯冷却模式，是当前核反应堆热工水力学研究的一个热点。反应堆主冷却剂系统出现中小破口时，堆芯内的热量主要由三种方式导出：蒸汽发生器(SG)的二次侧；破口流量热释放(早期)；应急堆芯冷却系统的再循环(中后期)。但是，如果反应堆冷却剂系统的破口尺寸大到一定的程度时，应急堆芯冷却系统注入的冷却剂不能使燃料元件完全淹没，堆芯上部裸露的燃料元件的温度就会升高甚至损坏。对于裸露的燃料元件来说，除了向上自然流动的少量蒸汽能够带走热量以外，堆芯燃料元件还有一种重要的冷却方式，即冷凝回流成为一种有效的热导出方法。蒸汽在流动过程中夹带有液体，当蒸汽通过狭窄流通或者受到冷却时，如流经堆芯上支撑板和蒸汽发生器入口时，就会发生蒸汽的冷凝回流。这些冷凝回流的冷却剂重新流回堆芯，会大大降低裸露燃料元件的温度，可使它们不会受到损坏。当蒸汽的流速大到一定程度时，就台发生回流极限(counter-current flow limitation，以下简称CCFL)。     

Mid-loop工况冷凝回流冷却措施分析

核电厂在Mid-loop工况下由正常余热排出（RHR）系统移出堆芯衰变热,一旦丧失RHR系统,若不采取措施,堆芯在沸腾后可能裸露并最终损坏。本工作以300 MW核电厂为对象,采用RELAP5/Mod3.2程序对Mid-loop工况下丧失RHR系统时的冷凝回流冷却措施进行分析。结果表明,在RCS回路封闭的情况下,两台蒸汽发生器（SG）均充满水,或1台SG充满水且辅助给水系统可用时,通过冷凝回流可维持24 h堆芯不裸露,即冷凝回流是可行的缓解措施之一。     

用FLICA3分析自然循环实验特性

ＦＬＩＣＡ３是由法国核研究中心发展的反应堆堆芯热工水力子通道分析研究，并在法国用于压水堆堆芯热工水力设计及安全分析。该文采用ＦＬＩＣＡ３程序对一个两项自然循环进行了分析计算并给出了有关结果。特别给出了堆芯内各段压降的程序计算值和实测值之间的比较。在计算中，通过调整有关参数，使计算结果和实验结果很好地吻合，获得了在不同的质量流速范围内计算空泡分额应该选取的分布系数和漂移速率。此外，还对ＦＬＩＣＡ不能     

全堆芯格林函数方法在秦山核电厂堆芯燃料管理中的应用

推导了全堆芯格林函数方法模型，用该模型发行了粗网差分堆芯燃料管理软件ＡＤＭＡＲＣ，形成快速堆芯燃料管理软件ＣＧＦＭＡＣ，并利用该软件对秦山核电厂的第１、２循环进行了堆芯计算。     

RELAP5／MOD2冷凝回流传热关系式分析

ＲＥＬＡＰ５／ＭＯＤ２冷凝回流传热关系式分析杨鲁伟陈听宽胡志宏徐进良（西安交通大学多相流国家重点实验室）关键词冷凝回流逆流流动极限层流冷凝紊流冷凝１前言ＲＥＬＡＰ５系列程序是美国爱达荷国家工程实验所（ＩＮＥＬ）发展起来的程序系列，主要用于核电厂系统在...     

GEDAK—生成INCORE—3D理论数据库的程序系统

应用综合法，以国内的堆芯燃料管理程序系统ＴＰＦＡＰ－ＣＦＭＰ为主体，研制了生成堆芯测量数据分析程序ＩＮＣＯＲＥ－３Ｄ所需的理论数据库计算程序系统ＧＥＤＡＫ。经秦山电厂若干测量数据的分析表明，该程序系统生成的数据库具有较高的计算精度。     

GEDAK──生成INCORE－3D理论数据库的程序系统

应用综合法，以国内的堆芯燃料管理程序系统ＴＰＦＡＰ－ＣＦＭＰ为主体，研制了生成堆芯测量数据分析程序ＩＮＣＯＲＥ－３Ｄ所需的理论数据库计算程序系统ＧＥＤＡＫ。经秦山核电厂若干测量数据的分析表明，该程序系统生成的理论数据库具有较高的计算精度。     

高功率研究堆低浓化物理特性研究

应用ＦＧ２ＤＢ两维两群扩散燃耗程序和带６９群中子截面库的ＣＥＬＬ栅元少群参数程序，对高功率研究堆低浓化堆芯进行了物理计算。ＬＥＵ燃料元件的铀密度为３．６－７．２ｇ／ｃｍ３，包壳厚度为０．３８－０．５６ｍｍ。结果表明：改变燃料芯体铀密度或厚度在物理上相当；各堆芯方案的控制棒价值等运行安全有关参数都可以接受。部分计算结果被拟合成线性或二次关系式以便于应用。给出了各堆芯的最小临界值、剩余反应性、运行寿期、快热中子通量和积分通量等物理参数。分析这些参数后指出：当Ｕ－２３５含量提高２０％或更多时，ＬＥＵ堆芯与ＨＥＵ堆芯的主要物理性能相近，这时快中子通量几乎不受影响，热中子通量的下降率近似正比于元件Ｕ－２３５含量增加率。但由于ＬＥＵ堆芯运行寿期的延长，对一般同位素生产与燃料元件辐照考验不会有明显影响。     

高通量工程试验堆80盒元件堆芯研究

通过理论分析和运行结果比较了高通量工程试验堆８０盒、６０盒工件堆芯性能。结果表明，ＨＴＦＴＲ８０盒元件堆芯在允许功率、材料辐照和单晶硅掺杂、钼锝同位素生产等方面与６０盒元件堆芯性能相同。８０盒元件堆芯更有利于５００ｋＷ回路入堆后堆的运行，有利于大幅度提高高比度＾６０Ｃｏ医疗源产量和元件利用率。     

蒸汽冷凝回流模型与应急堆芯冷却系统在中破口失水事故中的效益

在主回路冷段破口等效直径15.24cm的中破口失水事故分析，同时采用了不使用蒸汽冷凝回流模型、增大安注流量不使用蒸汽冷凝回流模型和使用蒸汽冷凝回流模型三种分析方法．分析结果表明：使用蒸汽冷凝回流模型时，回流的冷却剂可以有效地带走裸露燃料元件的热量，抑制燃料包壳温度升高．不使用蒸汽冷凝回流模型和增加安注流量时，裸露燃料元件的热量不能被带走，燃料包壳温度会升高．     

反应堆压力容器下降段水-蒸汽CCFL实验与模型研究

为探究反应堆压力容器下降段在喷放末期冷段安注过程中的水-蒸汽逆流特性,建立下降段逆向流动限制(CCFL)模型,开展了基于压力容器模化本体的下降段CCFL实验研究以及建模分析。通过实验研究获得了不同入口安注水流量、安注水过冷度、堆芯蒸汽流量等条件下的下降段环腔内的安注特性数据,并基于实验数据进行了CCFL建模分析。结果表明,开始发生CCFL的蒸汽无量纲流速与入口安注水无量纲流速呈现正相关,基于无量纲流速建立的模型斜率与入口安注水无量纲流速呈现高度指数关联。本文建立了适用于从不发生CCFL至不完全CCFL,再到完全CCFL的下降段水-蒸汽气液逆流全过程预测模型。     

Effect of Gas Atoms on Recovery of Work Hardening in Al-Li Alloys

In the first report of this study, dealing with CCFL and CCFL breakdown phenomena associated with the injection of emergency core cooling spray water into upper plenum during refill-reflood phase of a BWR LOCA, the following tests results were obtained.

The injected water maintained two-phase pool across the top of entire core after CCFL breakdown. The pool level oscillated near spray elevation. The objective of this paper is to clarify the mechanism of these phenomena, evaluating steam and spray flow effects on CCFL breakdown.

It is found that when spray flow rate was slightly larger than the CCFL drainage deter- mined by core steam flow, pool maintained at some constant level near spray elevation, after CCFL breakdown. On the other hand, when spray flow was appreciably larger than CCFL drainage, pool level slowly oscillated. The oscillation was caused by significant changes in steam condensation rate, and the corresponding subcooling penetration into the fuel bundles, when the pool level passed the spray elevation. The TRAC-BD1 analysis of test results suggested the small sector wall effect of test apparatus on CCFL breakdown phenomena.     

基于RELAP5的LOCA喷放阶段下降段内CCFL特性研究

反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL）,不利于应急冷却水顺利进入堆芯,极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为;通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性,并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明,由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性,随着蒸汽流量增大,在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大,较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系;在靠近破口的环路内注入的冷却水更难到达下腔室,而在远离破口环路的冷却水容易进入到下腔室;过热的蒸汽在流动过程中被冷却水冷却发生凝结现象,导致出口蒸汽流量小于进口蒸汽流量,且随着进口蒸汽流量的增大,凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。      

A review of CCFL phenomenon

Counter current flow limitation CCFL is an important phenomenon for numerous engineering applications and safety of light water reactors. In particular, the possible occurrence of CCFL in the hot-leg of a PWR during SBLOCA or LOCA accidents is of special interest for nuclear safety research. A review of the related literature has made in order to present the most important studies about the phenomenon and to reach common general understanding of the different factors that govern CCFL. Eventually this will allow explaining contradictions among different explanations provided by different authors. Most important factors were geometrical characteristics, liquid superficial velocity, and physical properties. The review shows that despite numerous experimental works, many scaling and geometrical effects are still not fully understood. For Instance there exist no consistent explanation of the channel diameter and inclined riser length effect upon results. The same can be stated-though to a minimum extent – for the inclination angle while channel length (or channel to diameter ratio) effect was clear and consistent. Since most experimental work was done in down-scaled hot-leg simulators, it becomes interesting to build a coherent knowledge about these effects and to explain arising contradictions in order to safely extrapolate results to full-scale hot-leg. The review has shown that many differences were simply due to geometrical effects, this leads to the need to “standardize” experimental data according to geometrical parameters. This should results in a better understanding of the phenomenon and corresponding scaling effects. Additionally, important variables such as: pressure drop, void fraction and shear stress were also investigated and discussed. A compilation of CCFL data was built and analyzed. Since the new simulation trends tend to use CFD codes where geometrical and spatial deviations are excluded by using 3D modeling, emphasis was placed upon introducing correlations for onset of CCFL out of collected data. Existing correlations for interfacial shear stress friction factor and the void fraction as a function of gas superficial velocity were also gathered and briefly discussed. The effect of condensation, physical parameters, and hysteresis upon CCFL was also introduced.     

Effect of steam condensation on countercurrent flow limiting in nearly horizontal two-phase flow

An analytical model that includes the steam condensation effect has been derived and a parametric study has been performed. In addition, a series of experiments were performed and a total of 34 experimental data for the onset of countercurrent flow limiting (CCFL) in nearly horizontal countercurrent two-phase flow have been obtained for various flow rates of water. Comparisons of the present CCFL data with slug formation models show that the agreement between the present as well as the existing model and the data is about the same. However, the deviation between Taitel and Dukler’s model predictions and the data is the largest when j_f<0.04 m s⁻¹. A parametric study of the effect of condensation using the present model shows that, when all local conditions are similar, the model predicted local gas velocities that cause the onset of flooding are slightly lower when condensation occurred. Based on the visual observation and the evaluation of the present work, it has been concluded that the criterion derived for the onset of slug flow can be directly used to predict the onset of inner flooding in nearly horizontal two-phase flow within the experimental ranges of the present work.     

Improved core design of the high temperature supercritical-pressure light water reactor

A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C.In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained.In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.     

Analyses of subcooled CCFL tests for evaluation of W̱COBRA/TRAC applicability

The W̱COBRA/TRAC-MOD7A, Rev. 1 code is currently licensed for best estimate large break LOCA analyses of 3 and 4 loop PWRs with emergency core cooling system injection located in the cold legs. As a part of the licensing effort to extend the code applicability to an upper plenum injection plant, the codes ability to predict subcooled flooding on a perforated plate was assessed by analyses of GE counter current flow limit tests and by comparison to the Bankoff flooding correlation (Bankoff, S.G., Tankin, R.S., Yuen, M.C., Hsieh, C., 1981). Counter current flow of air/water and steam/water through a horizontal perforated plate, Int. J. Heat Mass Transfer, 24 (8) 1382). The observed code model bias for subcooled CCFL can be eliminated by applying multiplication factors to the interfacial condensation and the interfacial drag models.     

RELAP5/MOD3.3 analysis of coolant depletion tests after safety injection failure during a large-break loss-of-coolant accident

This study is concerned with development of a coupled calculation methodology with which to continually and consistently analyze progression of an accident from the design-basis phase via core uncovery to core melting and relocation. Experiments were performed to investigate the core coolant inventory depletion after safety injection failure during a large-break loss-of-coolant accident in a cold leg utilizing the Seoul National University Facility (SNUF). The SNUF is an integral test loop scaled down to 1/6.4 in length and 1/178 in area from the Advanced Power Reactor 1400 MWe (APR1400). The SNUF tests are simulated with the RELAP5/MOD3.3 code. The test results revealed that the core coolant inventory decreased five times faster during the sweepout in the downcomer than after termination of the sweepout. The sweepout was observed to take place on top of spillover from the downcomer region to expedite the depletion of the core coolant inventory. The calculation results of RELAP5/MOD3.3 deviated from the experimental data in terms of entrainment from the surface of core coolant, condensation and sweepout in the downcomer. Thereby, the core coolant level was computed to decrease faster than the measured from the experiment due to the overestimated spillover by the evaporation of the entrained droplets by the uncovered heaters. Notwithstanding the occasional disparities, the code prediction is in reasonable agreement with the overall behavior of the tests.     

Applicability of Annular Flow Model to Counter-Current Flow in Debris Beds

Counter-current flow limitation (CCFL) is dominant phenomena for dryout in a debris bed which may be formed during a severe accident as observed in the Three-Mile Island unit-2. Actual CCFL situation in debris bed is very complex. It is difficult to treat the CCFL in the debris bed as it is. On the other hand, an annular flow model was developed to predict CCFL in a pipe by assuming a two-dimensional turbulent flow. If hypothetical flow channel were assumed for CCFL in the debris bed, CCFL in the debris bed could be treated with the same manner as for CCFL in a pipe. 'The purpose of this study is to investigate whether the annular flow model developed for CCFL in a pipe is applicable for CCFL in the debris bed or not. As the results, it is clarified that qualitative tendency of the CCFL in the debris bed consisting of larger particles than 3 mm is estimated by the annular flow model developed for CCFL in a pipe, although the difference between the calculation and the data is large in higher and lower gas velocity. It is also clarified that wall friction factor calculated with the present analysis is twice to forth larger than that in the single phase flow through porous media.     

The inherent boron dilution mechanism in pressurized water reactors

The purpose of this paper is to describe a mechanism that inherently causes boron dilution in pressurized water reactors (PWRs). The phenomenon is due to the fact that boric acid does not markedly dissolve into steam. This is relevant for transient and accident situations in PWRs where decay heat removal is accomplished by coolant vapourization and condensation, which inherently leads to formation of dilute plugs in the primary. In particular, it is found that inherent dilution will be inevitable for a range of small break loss of coolant accidents (SB LOCAs), with maximum amount of total diluted coolant mass exceeding 20 tonnes for a modern 1300 MWe PWR equipped with U-tube steam generators. A simple analysis of dilute plug motion during the late phases of a SB LOCA and core response to boron dilution shows that the damaging potential might extend to widespread fuel failures. Other transients and accidents are also discussed from the point of view of inherent dilution. Some possible remedies to the problem, as well as suggestions for further research, are presented.