先进压水堆事故缓解特性安全分析

发布时间：2023-05-27 03:02

　　作为第三代核电厂堆型,先进压水堆HPR1000具有非常好的固有安全性设计逻辑,并且通过在所有安全相关设计上贯彻众深防御原则来确保反应堆的安全。它拥有对设计基准事故后果的缓解措施,并包含对超设计基准事故的有效防御体系。本研究主要着眼于该堆型在二次侧非能动余热排除系统与VDA冷却能力上的改进。二次侧非能动余热排出系统是针对缓解全厂断电事故与辅助给水系统泵失效事故的先进设计,目的是利用非能动循环向蒸汽发生器二次侧供水。本研究以目前正在发展的堆型为研究对象,利用RELAP5/MOD3.4计算软件分析了 SGTR、LOFA、SBO的三种事故工况,评估这三种事故工况下核电厂关键系统的响应与重要参数的变化。本论文简述了对于这几种事故工况的主要缓解措施及设计特点。在SGTR事故中,并没有出现蒸汽发生器满溢工况,也没有其他的热工水力上限出现,这表明VDA被蒸汽发生器的水平所激励。在LOFA事故中,利用VDA泄出所提供的快速泄压可以提供一回路系统的最初非能动驱动力。非能动余热排出系统可以满足设计需要,并能够在全厂断电后72小时内成功进行余热排出功能。本文的结论对发展与HPR1000相似堆型的事故环节管理与...

【文章页数】：79 页

【学位级别】：硕士

【文章目录】：
摘要
ABSTRACT
CHAPTER 1. INTRODUCTIO
    1.1 BACKGROUND AND RESEARCH OBJECTIVES
        1.1.1 Background
        1.1.2 Research Objectives
    1.2 LITERATURE REVIEW
        1.2.1 Overview of Advanced Nuclear Power Plant
        1.2.2 HPR1000 NPP
    1.3 THESIS OUTLINE
CHAPTER 2. RESEARCH METHODOLOGY
    2.1 METHODOLOGY
    2.2 RELAP5 CODE
        2.2.1 Introduction
        2.2.2 Areas of application
        2.2.3 Modeling philosophy
CHAPTER 3. MODELING
    3.1 ADVANCED PRESSURIZED WATER REACTOR MODELING
    3.2 STEAM GENERATOR TUBE RUPTURE ACCIDENT MODELING
    3.3 PASSIVE RESIDUAL HEAT REMOVAL SYSTEM MODELING
CHAPTER 4. STEADY STATE AND TRANSIENT ANALYSIS
    4.1 STEAM GENERATOR TUBE RUPTURE ACCIDENT
        4.1.1 Overview of an SGTR accident
        4.1.2 Transient description and time sequence
        4.1.3 Steady state and transient analysis
    4.2 Loss of FLOW ACCIDENT
        4.2.1 Overview of loss of flow accident
        4.2.2 Transient description and time sequence
        4.2.3 Steady state and transient analysis
    4.3 STATION BLACK-OUT WITH PASSIVE RESIDUAL HEAT REMOVAL SYSTEMCOOLING CAPABILITY
        4.3.1 Overview of SBO and PRS
        4.3.2 Transient description and time sequence
        4.3.3 Steady state and transient analysis
CHAPTER 5. CONCLUSIONS AND FUTURE WORKS
    5.1 CONCLUSIONS
    5.2 FUTURE WORKS
REFERENCES
PAPERS PUBLISHED DURING THE MASTER'S DEGREE
ACKNOWLEDGEMENTS



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