先进核电厂半球顶安全壳抗震分析

安全壳是核电厂反应堆主厂房的围护结构,是防止设计事故发生时放射性物质扩散的最后一道屏障,是确保核电厂安全的关键设施.因此,必须在设计中考虑到安全壳在可能的、会引发重大核事故的意外荷载作用下的工作性能.地震是核电厂整个使用过程中有可能出现的自然灾害之一,并可能引发重大事故,所以,必须对安全壳结构进行严格的抗震性能分析,设计要保证预应力混凝土安全壳能够承受SSE作用而不被损坏.本文通过有限元模型的计算与分析,得到先进核电厂半球顶安全壳结构在SSE作用下的应力、变形、位移等地震反应,由此进行安全壳结构构件抗震分析计算.计算表明,半球顶安全壳结构在SSE作用下,安全壳结构安全可靠,结构的设计能够满足我国核电厂安全导则对抗震Ⅰ类结构的规定.     

接触爆炸荷载作用下核电站安全壳的动力响应分析

在现代战争中或遭到恐怖袭击时,核电站极有可能遭受精确制导装药的直接打击核电站安全壳是防止放射性物质向环境释放的大体积预应力混凝土筒壳结构,是特殊环境条件下的重点防护目标。根据核电站安全壳的结构形式．采用流固耦合算法,对装药接触核电站安全壳表面爆炸时,爆炸冲击作用下安全壳的动力响应进行了数值模拟,得出了装药接触爆炸时结构的破坏情况以及应力分布规律;数值模拟结果可为核电站在战时的安全防护对策制定提供参考依据。     

华龙一号安全壳热工响应确定论现实方法研究

失水事故（LOCA）是压水堆核电厂的一种典型设计基准事故,该事故后的安全壳热工响应过程,尤其是安全壳压力峰值直接影响安全壳结构的完整性。本文采用确定论现实方法（DRM）对华龙一号核电厂LOCA质能释放与安全壳热工响应进行分析研究。对关键参数进行敏感性分析及统计计算,并建立DRM惩罚模型。计算结果表明,DRM惩罚模型的计算结果始终高于95%置信水平下、95%概率下的统计计算值,DRM惩罚模型是保守的。DRM方法对于华龙一号核电厂的LOCA质能释放与安全壳热工响应分析是适用的。     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

CANDU6核电厂无过滤安全壳通风模式的研究

CANDU6核电厂早期设计未考虑严重事故对策,在严重事故下,CANDU6核电厂的安全壳容易失效。为了解决这一问题,本文研究了无过滤安全壳通风模式对CANDU6核电厂安全壳的影响。本文选取典型的全厂断电严重事故,利用重水蒸气回收系统作为无过滤安全壳通风的路径,初步研究了该通风模式下对安全壳完整性的保持和对裂变产物源项的滞留能力。研究表明:该通风模式可以有效保持安全壳的完整性,同时,对裂变产物源项也有一定的滞留能力。     

Analysis of long-term behaviour of nuclear reactor containment

For assessment of safety and durability of a nuclear power plant (NPP), knowledge of the containment behaviour under various service and extreme conditions is crucial. To perform reliable analysis of such a large-scale structure, a sufficiently realistic but still feasible numerical model must be used, in which the relevant physical phenomena are reflected. Therefore, a constitutive model for concrete including effects of moisture and heat transfer, cement hydration, creep, shrinkage and optionally microcracking of concrete should be chosen. The present paper focuses on the simulation of the service life of NPP containment, aiming to determine the material and model parameters to enable reliable prediction of structural behaviour under various conditions. The purpose of the work is to provide a numerical model calibrated using existing measurements to predict the long-term behaviour reliably. Extensive in situ measurements are used to calibrate the model and to check the validity of the model hypotheses. Moreover, the material model parameters are systematically re-calibrated based on the continuous monitoring of the structure. The structural integrity test is reanalysed numerically to show the model capability of predicting behaviour of the structure under given loading and climate conditions.     

先进压水堆核电厂氢气控制策略分析研究

新建核电厂的设计必须做到“实际消除”早期与大量放射性释放的可能性,氢气燃爆导致的安全壳失效是必须要“实际消除”的严重事故工况之一。因此对各种消氢措施的特点进行分析研究,建立联合消氢策略评价方法,可为先进压水堆核电厂氢气控制策略选择设计评价提供支持手段。根据严重事故管理中对氢气控制策略的考虑,研究安全壳内局部位置的可燃性是相关设计评价的关键问题。根据可燃性准则、火焰加速准则、燃爆转变准则,本文使用三维CFD程序对典型严重事故工况下安全壳蒸汽发生器隔间内的可燃性及氢气风险进行模拟分析。研究结果表明,虽然喷放源项中有大量水蒸气,蒸汽发生器隔间中仍有较大区域处于可燃限值以内,合理布置的点火器能在设计中点燃并消除氢气。本研究建立的分析方法能用于对核电厂氢气控制策略选择设计的评价。     

The IAEA accident management programme

Accident prevention and mitigation programmes and the Emergency Response System (ERS) are important elements of the Agency's activities in the area of nuclear power plant (NPP) safety. Safety Codes and Guides on siting, design, quality assurance and the operation of NPPs have been produced and are used by NPP operating organizations. Nuclear safety evaluation services are provided by the IAEA. The Emergency Response System and the International Nuclear Event Scale (INES) have been developed. The framework for the development of an accident management programme has been set up. The main goal is to develop an Accident Management Manual to provide a systematic, structured approach to the development and implementation of an accident management programme at NPPs. An outline of the Manual has been distributed and the first draft is available. The component parts are: co-ordinated research programmes (CRPs) on severe accident management and containment behaviour; the use of vulnerability analysis; mitigation of the effects of hydrogen, and generic symptom oriented emergency operating procedures. The IAEA provides guidance by the dissemination of information on methods for accident management; collates information on approaches in this field in different organizations and countries; and arranges exchange of experience and the promulgation of knowledge through the training of NPP managers and senior technical staff.     

Serviceability design load factors and reliability assessments for reinforced concrete containment structures

A reinforced concrete nuclear power plant containment structure is subjected to various random static and stochastic loads during its lifetime. Since these loads involve inherent randomness and other uncertainties, an appropriate probabilistic model for each load must be established in order to perform reliability analysis. The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops probability-based load factors for the limit state design of reinforced concrete containment structures. The purpose of constructing reinforced concrete containment structure is to protect against radioactive release, and so the use of a serviceability limit state against crack failure that can cause the emission of radioactive materials is suggested as a critical limit state for reinforced concrete containment structures. Load factors for the design of reinforced concrete containment structures are proposed and carried out the reliability assessments.     

对钠冷快堆安全壳设计的探讨

安全壳是在核电厂纵深防御原则中的最后一道屏障,是保证核电厂在严重事故条件下安全的重要设施.由于钠冷快堆的固有安全性,其安全壳与压水堆安全壳的功能和作用相同.通过对钠冷快堆安全壳的设计分析,阐述了其在设计基准、结构设计上区别于压水堆安全壳的特点,为确定钠冷快堆安全壳设计和审评原则做出了有益的探索.     

地下核电站安全壳再循环系统设计的初步论证

核电站建造于地下,反应堆厂房洞室外具备天然的裂变产物屏障,在安全壳外洞室内设置安全壳再循环系统,预防并缓解放射性裂变产物释放,维持安全壳的完整性。该系统同时整合了卸压、过滤、排热安全功能,充分发挥地下核电站重力补水和天然屏障的安全优势,可以非能动运行。本文通过简单的计算分析开展初步论证,证明该系统可以有效实现三大安全功能,是适合于地下核电站的安全系统。     

考虑一致危险性谱的核电厂房结构振动台试验研究

为得到适合特定核电厂所需要的反应谱,考虑具体的场地条件及地震动参数,采用随机模拟方法与概率危险性分析相结合的方式,建立了生成超越概率为10-4的一致危险性谱（UHS）的方法。为进一步研究核电结构的抗震性能及UHS在实际核电结构中的适用性,设计和制作了1∶20的核电厂房结构模型进行振动台试验,采用2条天然波及UHS、厂址谱（SL-2）、RG1.60谱所生成的人工波对结构的响应进行比对分析。结果表明,不同地震波对核电结构的响应有所差异,UHS生成的人工波对上部结构加速度放大效应以及位移影响较大,对应的楼层反应谱幅值相对其他反应谱较高,进行结构及设备抗震设计时应予以考虑。      

基于PSA技术的核电厂数字化仪控系统可靠性设计及应用

传统意义上核电厂数字化仪控系统主要依靠提升设备的可靠性来满足电厂安全目标。随着监管要求的逐步提高,在提升设备可靠性基础上,基于概率论技术的设计手段逐步成为核电厂安全设计新的研究方向。本文应用概率安全评价（PSA）技术,对典型电厂始发事件进行分析及研究,之后对仪控设计方案整体进行PSA建模,再将其置于电厂PSA模型中,通过定量评估分析,识别薄弱环节,给出优化改进措施。在此基础上提出了一套确保核电厂仪控系统满足整体安全目标的可靠性设计流程。      

GOTHIC程序在海洋核动力平台反应堆舱安全分析领域的适用性评估

针对海洋核动力平台反应堆舱热工水力分析程序缺乏的现状,以一回路失水事故（LOCA）下反应堆舱压力响应为评价基准,基于安全壳现象识别与排序表（PIRT）分析方法,通过开展LOCA下反应堆舱热工水力现象识别、现象分级研究,建立了反应堆舱PIRT。通过开展GOTHIC程序模型验证矩阵与PIRT的匹配性分析,确认GOTHIC程序在海洋核动力平台反应堆舱热工水力分析领域的适用性。本文分析方法对其他安全分析程序在核电等领域的跨领域适用性评估具有一定参考价值。      

On the failure probability of the containment under accidental aircraft impact

Reliability concepts are applied to the containment of nuclear power plants (NPP) under accidental aircraft impact. The statistical properties of the various input variables effecting the loading condition such as mass, stiffness, impact velocity, etc. are discussed. The loading time history is modeled by a nonstationary random process. As aircraft engines detach immediately after impact, the analysis of the hard engine and the soft fuselage impact has to be carried out separately. The present analysis shows that the failure probability of the containment due to fuselage impact is negligibly small, i.e. < 10⁻⁸.     

福岛核事故后核安全改进行动及安全要求研究

2011年“3·11”日本福岛第一核电厂严重核事故给世界核工业界造成了巨大影响。本文总结了从福岛核事故中汲取的教训,介绍了主要核电国家,如美国、日本、法国以及中国在福岛核事故后十年来实施的一系列核安全改进行动和核安全法规标准修订。阐述了核安全要求和核安全理念在中国的实施现状及实践,包括实际消除早期或大量放射性释放、事故工况划分、纵深防御概念、移动设施配置等,对后续核安全发展方向进行探讨并提出建议。     

RTDP和MSG方法比较研究

偏离泡核沸腾比(DNBR)设计限值是反应堆安全分析的基础,合理的计算方法有助于发掘更多的安全裕量。本文对比分析了核电站设计中常采用的RTDP和MSG方法,介绍了方法原理和计算流程,并以广东岭澳一期核电站为例,分别进行了DNBR设计限值计算。结果表明:虽然两种方法的原理不同,但在相同的工况和统计学输入参数条件下,DNBR设计限值相近,设计中可根据具体应用需求进行选择。     

核电厂数字化安全系统人机接口设计研究

核电厂安全系统人机接口分别与电厂安全系统和整个仪表与控制(I&C)系统人机接口相关。本文对核电厂控制室中数字化安全系统人机接口的设计进行了描述,同时也论述了作为安全系统重要组成部分的反应堆保护系统人机接口的有关设计内容以及在安全系统人机接口设计中应关注的有关要求,并展望了未来在新技术方面的应用发展趋势。     

Feasibility of underground NPP and its MCDA-based optimization

The underground nuclear power plant(NPP)makes full use of land resources, reduces costs, makes better use of its passive safety, and avoids radioactivity release into the atmosphere in serious nuclear accidents.In this paper, for obtaining comprehensive and integrated analyses on this new NPP design, we introduce four kinds of underground NPP designs, analyze the feasibility of each design from various aspects, and use the multiple criteria decision analysis method to choose the best option.     

Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant,WWER-1000/320

This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.