先進壓水堆事故緩解特性安全分析
發(fā)布時間:2023-05-27 03:02
作為第三代核電廠堆型,先進壓水堆HPR1000具有非常好的固有安全性設計邏輯,并且通過在所有安全相關設計上貫徹眾深防御原則來確保反應堆的安全。它擁有對設計基準事故后果的緩解措施,并包含對超設計基準事故的有效防御體系。本研究主要著眼于該堆型在二次側(cè)非能動余熱排除系統(tǒng)與VDA冷卻能力上的改進。二次側(cè)非能動余熱排出系統(tǒng)是針對緩解全廠斷電事故與輔助給水系統(tǒng)泵失效事故的先進設計,目的是利用非能動循環(huán)向蒸汽發(fā)生器二次側(cè)供水。本研究以目前正在發(fā)展的堆型為研究對象,利用RELAP5/MOD3.4計算軟件分析了 SGTR、LOFA、SBO的三種事故工況,評估這三種事故工況下核電廠關鍵系統(tǒng)的響應與重要參數(shù)的變化。本論文簡述了對于這幾種事故工況的主要緩解措施及設計特點。在SGTR事故中,并沒有出現(xiàn)蒸汽發(fā)生器滿溢工況,也沒有其他的熱工水力上限出現(xiàn),這表明VDA被蒸汽發(fā)生器的水平所激勵。在LOFA事故中,利用VDA泄出所提供的快速泄壓可以提供一回路系統(tǒng)的最初非能動驅(qū)動力。非能動余熱排出系統(tǒng)可以滿足設計需要,并能夠在全廠斷電后72小時內(nèi)成功進行余熱排出功能。本文的結論對發(fā)展與HPR1000相似堆型的事故環(huán)節(jié)管理與...
【文章頁數(shù)】:79 頁
【學位級別】:碩士
【文章目錄】:
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
本文編號:3823763
【文章頁數(shù)】:79 頁
【學位級別】:碩士
【文章目錄】:
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
本文編號:3823763
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