铅铋堆蒸汽发生器传热管破裂事故三维程序开发及验证

蒸汽发生器（SG）传热管破裂事故（SGTR）是铅铋堆设计必须重点考虑的安全问题之一。针对铅铋堆SGTR，为解决其复杂结构环境中压力波的三维传播与蒸汽的三维迁移难题，基于多相流欧拉流体动力学理论，开展了“铅铋-水”相互作用三维数值模型与算法研究，研制了专用程序，并采用实验对比和程序对比技术手段进行了程序验证，验证结果符合较好。研究结果表明：对于描述铅铋堆SGTR过程中“铅铋-水”相互作用行为，本文采用的相关数值理论与模型具有较好的适用性；对于研究复杂结构环境下铅铋堆SGTR的三维演化现象，包括压力波传播、蒸汽迁移，本文所开发的三维程序具有重要的潜在应用价值。本文研究成果有望为我国铅铋堆SGTR分析提供有力支撑。     

铅铋快堆SGTR事故下高压过冷水注入高温铅铋合金流动传热数值模拟研究

铅铋快堆内蒸汽发生器传热管两侧为高压过冷水和高温铅铋冷却剂，传热管两侧较大的压差和温差以及液态铅铋合金（LBE）的腐蚀效应可能造成蒸汽发生器传热管破裂（SGTR）事故。深入研究事故后高压过冷水冲击高温液态LBE的射流沸腾和相变产物蒸汽扩散的特征，具有十分重要的学术意义和工程应用价值。为揭示事故工况下液态LBE与水相互作用的传热传质机理，基于流体体积（VOF）方法，结合LES湍流模型和Lee相变模型，建立了水/蒸汽-液态铅铋多相流动与传热的三维数值计算模型，系统研究了高压过冷水注入高温LBE内发生的相变传热过程。结合注入压力及过冷水温度等因素，分析了射流沸腾过程中不同工况对射流形态、迁移深度以及沸腾行为的影响，研究结果可为SGTR事故工况下堆芯安全性预测提供指导。     

燃料组件堵流工况下铅铋-氩气两相流的传热压降特性分析

带绕丝燃料组件的堵流事故是铅冷快堆安全分析的重要工况之一。由于在铅铋自由液面处的气体夹带或在气体增强自然循环条件下存在铅铋-氩气的两相流情况,可能引起燃料组件堵流工况下的局部热工水力特性变化。本文通过计算流体力学软件Fluent,对带绕丝19棒束燃料组件进行建模,模拟分析了堵流工况下的铅铋-氩气两相流传热压降特性,并对两相流模型进行了对比验证,对入口雷诺数、堵块孔隙率、氩气气泡直径等因素进行参数敏感性分析。结果表明：在堵流条件下氩气气泡的流动行为包括逃逸、耗散和受限,在气相体积分率较高的区域会产生局部微正压及过热现象。研究结果可为铅冷快堆堵流事故的安全分析提供参考。     

铅水反应中铅铋合金凝固的数值模拟

为了研究铅铋合金在蒸汽发生器传热管破裂（SGTR）事故所引发的铅铋合金与水反应过程中的凝固机理,通过耦合VOF模型、Realizable k-ε湍流模型、凝固传热模型,利用FLUENT软件建立了铅铋合金与水反应过程的二维仿真模型,并将该模型与现有反应实验的结果进行对比验证。随后基于热焓法建立可以直观描述铅铋合金凝固现象的凝固传热特性热焓方程,通过控制模型变量研究影响铅铋合金凝固发生的因素及条件,最后将该模型应用于复杂结构场景中。结果表明,铅铋合金与水的温差、水流喷射初始速度、注水管径是影响铅铋合金凝固的主导因素,本文提出的模型具有较高可靠性,能够模拟实际工况中铅铋合金的凝固现象。本研究所得到的机理性结论与现象学结论能够为铅基快堆安全分析提供理论支撑。      

民用小堆SGTR防满溢设计改进

民用小堆因单位功率下的蒸汽发生器（SG）汽空间偏小,稳压器容积和SG传热管内径偏大等特点,会引起蒸汽发生器传热管破裂（SGTR）事故快速满溢。本文采用RELAP5程序对民用小堆SGTR事故开展了优化措施研究,并提出极限单一故障下防止SG发生满溢的工程可行方案,即增加SG高水位排放液体的溢流管线或提高二次侧设计压力且同时增加自动的安注闭锁信号,保证在事故过程中蒸汽发生器不满溢和放射性排放满足限值要求。在民用小堆专设设备基本不变的前提下,针对系统进行了优化,极大地提升了安全性,为民用小堆设计改进提出了工程可行方案。     

SGTR事故下气泡穿透深度模拟实验

蒸汽发生器传热管破裂(SGTR)后,气泡在冷却剂中的穿透深度影响铅基冷却反应堆的安全运行。针对中国铅基反应堆SGTR事故,实验营造不同气体泄漏量,利用高速摄影技术对气泡在水介质中的穿透深度特性进行了模拟实验研究。观察了气泡流动流型演化全过程,得到了气泡流型及穿透深度的初步实验数据,并推导出气泡无量纲穿透深度与弗劳德数间的准则关系式,在弗劳德相似准则基础上该关系式可应用于密度比小的气泡在液态金属冷却剂中的注入过程。实验结果表明,在破口面积一定的条件下,气泡穿透深度与气体初始速度呈正比。由量纲分析得到气泡穿透深度关系式与文献的实验结果吻合较好。     

基于CFD的池式快堆多物理耦合分析方法研究

基于临界/次临界点堆中子动力学模型、燃料棒传热模型、热交换器和多孔介质等辅助热工水力模型，采用显式迭代和动态链接库技术（DLL），利用商用计算流体力学（CFD）程序FLUENT的用户自定义函数（UDF）实现中子动力学、燃料棒热传导等和快堆堆池冷却剂流动换热的耦合计算，开发池式快堆多物理耦合计算程序CFD/PF。采用CFD/PF开展小型自然循环铅铋快堆SNCLFR-10无保护超功率事故（UTOP）模拟，并与国际知名快堆多物理耦合分析程序SIMMR-III的计算结果开展Code-to-Code对比分析。研究结果表明:CFD/PF与SIMMER-III的分析结果吻合良好，耦合程序的开发取得了初步成功，可用于分析池式快堆堆池内的复杂三维流动和换热现象。      

核电厂SGTR规程优化研究

核电厂通过应急运行规程（EOP）来缓解蒸汽发生器传热管破裂（SGTR）事故,SGTR事故分析结果显示,在缓解过程中操纵员开启稳压器卸压阀进行反应堆冷却剂系统（RCS）降压后,安全注射（简称“安注”）流量大幅增加,导致稳压器水位大幅增加,可能存在潜在的危险。本文目的是为了更好地缓解SGTR事故,使事故缓解过程中稳压器水位不致上升过高,确保核电厂安全。通过对EOP缓解步骤进行优化,提前切除一列安注,并对优化后的EOP缓解事故过程进行分析计算,最终结果显示稳压器最高水位下降,减少了稳压器水位过高的风险,为后续核电厂规程的改进提供了依据。      

直流蒸汽发生器传热管破裂事故分析

通过对直流蒸汽发生器传热管破裂（SGTR）事故的分析,可看出RELAP5瞬态分析程序能较好地模拟一体化反应堆在SGTR事故后的事件响应序列及主要热工水力现象,例如环路的不对称效应、主回路的自然循环等。一体化反应堆在发生SGTR事故后,可通过一系列安全与保护系统的动作得到有效缓解,并最终能应用非能动余热排出系统（PRHRS）的自然循环导出堆芯余热,使反应堆处于安全状态。同时,受事故影响蒸汽发生器压力在PRHRS投入运行后会快速升高,最终与一回路压力相平衡,此后,破口处的泄漏也会终止。此外,本文还研究了破口处临界流量及其积分流量结果不确定性的影响因素,其中主要考虑了采用不同的临界流模型和破口建模方式等两个方面。     

基于EOS压力迭代的压力波传播CFD模型研究

核能系统内压力波传播将造成水力学载荷效应，实现对压力波传播过程的精确模拟对结构载荷、应力分析而言尤为重要。一维系统分析程序（RELAP5、TRACE等）可用于水堆（压水堆、沸水堆）压力波传播模拟、分析。然而，对于一体化池式堆系统，聚变堆液态金属包层出现的压力波传播现象，如铅铋反应堆蒸汽发生器传热管破裂事故引发的压力波传播问题，系统分析程序不够精细，无法揭示复杂结构空间内的压力波传播行为。针对此问题，本文提出了一种基于EOS压力迭代的二维压力波传播CFD模型及相关算法，编制了程序代码。采用一维空气激波管基准例题和二维压力波传播实例进行了程序验证。前者与理论解和ANSYSFluent对比，后者与ANSYS Fluent对比。验证结果表明本文提出的数值方法、模型可较为合理、准确地模拟单相压力波的二维传播现象。     

Experimental study on gas-injection enhanced circulation performed with the CIRCE facility

This paper describes the results of an experimental campaign concerning the possibility of achieving a steady state circulation by gas-injection in a pool containing lead–bismuth eutectic (LBE) as working fluid. The activity was aimed at gaining information about the basic mechanisms of the gas injection enhanced circulation intended as a pumping system for a liquid metal cooled reactor. In particular, the paper is focused on the experimental work performed in the CIRCE large-scale facility, installed at the ENEA Brasimone Centre for studying the fluid-dynamic and operating behaviour of ADS reactor plants cooled by LBE. The gas enhanced circulation tests were carried out for different LBE temperatures (from 200 to 320 °C), under isothermal conditions and with a wide range of argon injected flow rates (from 0.5 to 7.0 Nl/s). The gas is injected from the bottom of the riser, by means of an appropriate nozzle, and the liquid metal flow rate is measured by a Venturi-Nozzle flow meter installed in the single phase part of the test section. The obtained results allowed formulating a characteristic curve of the system and evaluating the void fraction distribution along the riser path by means differential pressure measurements, which play an important role to generating the driving force for the circulation.     

Addressing the heavy liquid metal – Water interaction issue in LBE system

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.     

An analysis of containment responses during a station blackout accident

An analysis of the responses of the containment during a station blackout accident is performed for the APR1400 nuclear power plant using MELCOR 2.1. The analysis results show that the containment failure occurs at about 84.14 h. Prior to the failure of the reactor vessel, the containment pressure increases slowly. Then, a rapid increase of the containment pressure occurs when a large amount of hot molten corium is discharged from the reactor pressure vessel to the cavity. The molten corium concrete interaction (MCCI) is arrested when water is flooded over a molten corium in the cavity. The boiling of water in the cavity causes a fast increase in the containment pressure. During the early phase of the accident, a large amount of steam is condensed inside the containment due to the presence of the heat structures. This results in a mitigation of a containment pressure increase. During the late phase, the containment pressure increases gradually due to the addition of steam and gases from an MCCI and water evaporation. It was found that two-thirds of the total mass of steam and gases in the containment is from an MCCI and one-third of the mass is from water evaporation.     

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

During a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.     

Code development and safety analyses for Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR) can produce steam from the direct contact of feed-water and lead bismuth eutectic (LBE) in the chimney of 3 m height, which eliminates the bulky and flimsy steam generators. Moreover, as the coolant LBE is driven by the buoyancy of steam bubbles, the primary pump is not necessary in the reactor. The conceptual design makes the reactor simple, compact and economical. Owing to the large thermal expansion coefficient of LBE and good performance of steam lift pump, the reactor is expected to have good passive safety. A new computer code is developed to investigate the thermal–hydraulic behaviors and safety performance of PBWFR in the present work. Unprotected rod run-out transient over power (UTOP) and unprotected loss of flow (ULOF)/unprotected loss of heat sink (ULOHS) are simulated to test and verify its safety. The results show that PBWFR has very good inherent safety due to the satisfactory neutron and thermal–physical properties of LBE. Cladding materials turn to be the key factor to restrict its safety performance and UTOP is more dangerous for PBWFR. It's suggested that it should appropriately reduce the maximum value of the control rods to mitigate the consequence of UTOP due to good reactivity feedbacks in the core.     

基于扩散界面法的液态铅铋合金中气泡上升行为模拟

基于扩散界面法,对单个氮气气泡在液态铅铋合金内从静止到充分发展整个过程中的动力学行为进行数值模拟,得到气泡形变特性和气泡上升速度随时间的变化关系,将模拟结果与Grace经验关系图对比,发现模拟得到的气泡形变结果在Grace经验关系图中均可找到且很好地吻合,从而验证了扩散界面法在模拟液态铅铋合金中气泡上升行为的可行性和准确性。同时基于界面扩散法的模拟,对比了5种不同初始直径的氮气泡在液态铅铋合金中的上升行为,发现初始直径较小的气泡在上升过程中扰动会更剧烈,初始直径较大的气泡在上升过程中易发生分裂现象。     

铅铋快堆螺旋管直流蒸汽发生器热工水力特性数值研究

本研究以铅铋快堆螺旋管直流蒸汽发生器（HOTSG）设计结构为研究对象,采用精细网格与多孔介质相结合的物理建模方法,通过一次侧三维湍流计算与二次侧用户自定义函数（UDF）分区传热计算相耦合的手段,在FLUENT求解器中开展了蒸汽发生器的热工水力特性数值分析研究。研究表明:铅铋入口附近的流量分配孔和腔室对应的直管段区域出现铅铋流速峰值,径向最大速度为0.431 m/s;入口腔室至管束区位置受到阻力突变的影响,压力、横流速度、轴向速度变化较大;热工参数变化符合流动与传热机理,临界热流密度（CHF）点附近一二次侧温差最大为109.61 K,此处最大热流密度为323.55 kW/m2。该研究将为铅铋快堆HOTSG结构设计、流致振动及安全评价提供重要的参考。      

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.     

Estimation of ex-vessel steam explosion pressure loads

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel–coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel–coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.     

Acoustics impact on steam dryer induced by main steam line break in a boiling water reactor

In boiling water reactor (BWR) design, significant acoustic pressure loads impact the steam dryer hood as a result of the main steam line break outside containment (MSLB) event. When a main steam line breaks, it is assumed that the pipe instantaneously breaks completely open to the ambient environment (double-ended guillotine break). Due to the huge pressure difference between the inside of the reactor pressure vessel (RPV) and surrounding ambient environment, a shock wave will form at the break point and burst into the surrounding environment. At the same time, an expansion wave will travel upstream through the main steam line to the RPV, which results in a pressure reduction on the outside of the steam dryer hood. This expansion wave will create a substantial pressure difference between the two sides of the steam dryer hood with a resultant high stress on the hood. This differential pressure load is the acoustic load used in the structure design evaluations for this event. A key design basis requirement for the steam dryer is to maintain structural integrity during transient, and accident conditions. Demonstration that the steam dryers meet this design basis requires a calculation of the magnitude of the acoustic load on the steam dryer during a MSLB. In this study, computational fluid dynamics (CFD) is used as an alternate calculation method to investigate the phenomenon of MSLB. Transient simulations with fine time steps were carried out. The results show that CFD is a useful tool to provide additional information on the acoustic load as compared to the traditional methods. From the CFD results, the minimum pressure value and its distribution area at different flow times was identified. Through the modeling, an understanding of the detailed transient flow field, particularly the acoustic pressure field near the dryer hood during the MSLB was achieved.