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均匀碎片床与分层碎片床干涸现象及热流密度实验研究

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核动力工程 2024年第1期45卷 60-64页

作者：邹文斌佟立丽曹学武上海交通大学机械与动力工程学院上海200240

碎片床的可冷却性是反应堆严重事故缓解策略关注的问题,碎片床结构对其可冷却性有重要影响。本文搭建了体积加热碎片床可冷却性实验装置,开展了均匀碎片床、轴向分层碎片床、径向分层碎片床的可冷却性实验,揭示了不同类型碎片床的干涸特... 详细信息

碎片床的可冷却性是反应堆严重事故缓解策略关注的问题,碎片床结构对其可冷却性有重要影响。本文搭建了体积加热碎片床可冷却性实验装置,开展了均匀碎片床、轴向分层碎片床、径向分层碎片床的可冷却性实验,揭示了不同类型碎片床的干涸特性,获得了干涸热流密度(DHF)。研究结果表明:均匀碎片床条件下,小颗粒较大的气液流动阻力使得渗透率下降,导致小颗粒均匀碎片床DHF较小;轴向分层碎片床条件下,由于分层界面孔隙率降低引起的较大阻力,导致轴向分层碎片床DHF远小于顶部小颗粒组成的均匀碎片床DHF;径向分层碎片床条件下,由于小颗粒层较大的气液流动阻力会引起蒸汽向大颗粒层迁移和聚集,导致DHF低于大颗粒组成均匀床的DHF。

关键词：严重事故分层碎片床可冷却性干涸热流密度(DHF)

基于多孔介质方法的钠冷快堆冷却剂沸腾现象模拟

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强激光与粒子束 2024年第7期36卷 106-114页

作者：惠天宇佟立丽曹学武上海交通大学机械与动力工程学院上海200240

基于两流体六方程模型针对钠的气液两相分别构建守恒方程,采用蒸发冷凝模型表征两相质量交换,分别使用显式和隐式处理方法对蒸发冷凝模型进行计算,同时考虑了Sobolev阻力模型、两相对流换热模型以及相间动量交换等本构关系,开发了适用... 详细信息

基于两流体六方程模型针对钠的气液两相分别构建守恒方程,采用蒸发冷凝模型表征两相质量交换,分别使用显式和隐式处理方法对蒸发冷凝模型进行计算,同时考虑了Sobolev阻力模型、两相对流换热模型以及相间动量交换等本构关系,开发了适用于模拟钠冷快堆冷却剂沸腾的多孔介质分析方法,利用KNS-37失流实验L22工况数据进行了对比验证,并利用L29工况流量数据验证模型的适用性。结果表明,所建立的钠沸腾多孔介质分析方法可以较好地模拟钠冷快堆沸腾现象,预测沸腾发生时间在6.3 s左右,与实验相差0.2 s,温度和流量的总体变化趋势与实验数据吻合较好。

关键词：钠冷快堆钠沸腾多孔介质方法两流体六方程

Analysis of hot leg natural circulation under station blackout severe accident

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Nuclear Science and Techniques 2007年第2期18卷 123-128页

作者： DENG Jian cao xuewu School of Nuclear Science and Engineering Shanghai Jiaotong University Shanghai 200240 China

Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor (PWR). In a station blackout accident with no recovery of steam generator (SG) auxiliary feedwater (TMLB' severe accident scenario), the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg, and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg. The recirculation ratio and the hot mixing factor are also calculated and discussed.

关键词：压水堆全场断电严重事故热管段自然循环

Methodology for reliability allocation based on fault tree analysis and dualistic contrast

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Nuclear Science and Techniques 2008年第4期19卷 251-256页

作者： TONG Lili cao xuewu School of Nuclear Science and Engineering Shanghai Jiao Tong University Shanghai200240 China

Reliability allocation is a difficult multi-objective optimization problem. This paper presents a methodology for reliability allocation that can be applied to determine the reliability characteristics of reactor systems or subsystems. The dualistic contrast,known as one of the most powerful tools for optimization problems,is applied to the reliability allocation model of a typical system in this article. And the fault tree analysis,deemed to be one of the effective methods of reliability analysis,is also adopted. Thus a failure rate allocation model based on the fault tree analysis and dualistic contrast is achieved. An application on the emergency diesel generator in the nuclear power plant is given to illustrate the proposed method.

关键词：反应堆安全事故维修方法核技术

Effect of water injection on hydrogen generation during severe accident in PWR

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Nuclear Science and Techniques 2009年第5期20卷 312-316页

作者： TAO Jun cao xuewu School of Mechanical and Power Engineering Shanghai Jiao Tong University Shanghai 200240 China

Effect of water injection on hydrogen generation during severe accident in a 1000 MWe pressurized water reactor was studied. The analyses were carried out with different water injection rates at different core damage stages. The core can be quenched and accident progression can be terminated by water injection at the time before cohesive core debris is formed at lower core region. Hydrogen generation rate decreases with water injection into the core at the peak core temperature of 1700 K, because the core is quenched and reflooded quickly. The water injection at the peak core temperature of 1900 K, the hydrogen generation rate increases at low injection rates of the water, as the core is quenched slowly and the core remains in uncovered condition at high temperatures for a longer time than the situation of high injection rate. At peak core temperature of 2100–2300 K, the Hydrogen generation rate increases by water injection because of the steam serving to the high temperature steam-starved core. Hydrogen generation rate increases significantly after water injection into the core at peak core temperature of 2500 K because of the steam serving to the relocating Zr-U-O mixture. Almost no hydrogen generation can be seen in base case after formation of the molten pool at the lower core region. However, hydrogen is generated if water is injected into the molten pool, because steam serves to the crust supporting the molten pool. Reactor coolant system (RCS) depressurization by opening power operated relief valves has important effect on hydrogen generation. Special attention should be paid to hydrogen generation enhancement caused by RCS depressurization.

关键词： PWR 核技术研究发展 RCS

Hydrogen and steam distribution following a small-break LOCA in large dry containment

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Nuclear Science and Techniques 2007年第3期18卷 181-185页

作者： DENG Jian cao xuewu School of Nuclear Science and Engineering Shanghai Jiaotong University Shanghai 200240 China

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations. Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident (LOCA) without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investi- gated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump (RCP) compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accu- mulation.

关键词：氢蒸发严重事故氢分布小破口事故

Simulation of the mixing process in FCIs with hydrodynamic fragmentation model

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Nuclear Science and Techniques 2007年第4期18卷 242-246页

作者： LIN Qian cao xuewu School of Nuclear Science and Engineering Shanghai Jiaotong University Shanghai 200240 China

Fuel Coolant Interactions (FCIs) are important issues in nuclear reactor severe accident analysis. In FCIs, fragmentation model of molten droplets is a key factor to estimate degree of possible damage. In this paper, the mixing process in FCIs is studied by the simulation of MIXA experiment with hydrodynamic fragmentation model. The result shows that hydrodynamic fragmentation model underestimates the fragmentation rate of high temperature molten droplets under the condition of low Weber numbers. It is concluded that models based on thermal fragmentation mechanism should be adopted to analyze the FCI process and its consequence.

关键词：核反应堆燃料冷冻作用事故燃料熔化

Simulation of quenching in horizontal geometry channel

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Nuclear Science and Techniques 2010年第6期21卷 370-374页

作者： Yuan Jingtian Tong Lili cao xuewu Shanghai Jiao Tong Univ Sch Mech Engn Shanghai 200240 Peoples R China

Two tests of refilling and reflooding in the horizontal geometry channel under different pressure conditions,respectively,are conducted to understand the thermal hydraulics behavior during a LOCA in the CANDU *** case of refilling a hot horizontal tube test and four cases of reflooding a horizontal fuel channel test series (35%and 60%break size of reactor inlet header with single-end injection or both-end injection,respectively) have been *** horizontal channel is simulated with the model splitting as parallel pipes with cross flow junction *** simulation results have been compared with the tests in terms of wall temperature or heated-pin surface *** results are reasonably acceptable and it shows that the code is applicable on horizontal channel quenching.

关键词：水平测试通道测试模型模拟几何淬火 CANDU堆压力条件热工水力

Analysis of severe core damage accident progression for the heavy water reactor

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Nuclear Science and Techniques 2010年第4期21卷 251-256页

作者： TONG Lili YUAN Kai YUAN Jingtian cao xuewu School of Mechanical Engineering Shanghai Jiao Tong University' Shanghai 200240 China

In this study,the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis *** selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety *** progressions of severe accident included a set of failed safety systems normally operated at full power,and initiative events led to primary heat transport system inventory blow-down or boil *** core heat-up and melting,steam generator response,fuel channel and calandria vessel failure were *** results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault.

关键词：重水反应堆损坏事故蒸汽发生器堆芯 CANDU堆严重事故热传输系统事故分析