利用TRACE進行沸水式核能電廠於ATWS下MSIV關閉後之抑壓池暫態分析

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研究生：	蔡斐然
論文名稱：	利用TRACE進行沸水式核能電廠於ATWS下MSIV關閉後之抑壓池暫態分析 Use Trace Code to Analyze ABWR/BWR Suppression Pool Transient Under ATWS'S MSIV Closure Situation
指導教授：	施純寬王仲容
口試委員:
學位類別：	碩士 Master
系所名稱：	原子科學院 - 工程與系統科學系 Department of Engineering and System Science
論文出版年：	2008
畢業學年度：	96
語文別：	英文
論文頁數：	150
中文關鍵詞：	核四、龍門電廠、主蒸汽隔離閥、抑壓池
外文關鍵詞：	TRACE, ABWR, MSIV
相關次數：	點閱：2 下載：0
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摘要
本論文以核一、核二、核四廠之抑壓池做為研究對象，以TRACE為模擬程式，針對100% POWER主蒸汽隔離閥關閉之ATWS安全事故發生時，安全釋壓閥(SRV)開啟所排放的蒸汽進入液壓池的水中冷卻高溫高壓蒸汽之狀態做分析，在以TRACE為程式架構前提下，使用CONTAN模組建立抑壓池系統，作為BWR/ABWR抑壓池暫態反應分析。
在此探討之反應爐預期暫態未急停是指核能電廠發生預期運轉事件達反應爐保護系統動作設定，但動作失效無法使反應爐急停，此處所指之預期運轉事件在核能電廠運轉年限中，預期會發生一次或多次。10CFR50.62[9] 對預期暫態未急停之要求為：增設一套再循環水泵跳脫系統、增設一套替代性插棒系統、增加硼酸濃縮度、增加一套自動注硼系統、及增加一套飼水自動回退系統。
TRACE系統模式驗證工作以核一、核二、核四廠設計資料與ATWS事故分析報告為基礎，藉由這三座核電廠事故案例以TRACE做為驗證流程流量與壓降計算模式，並以雷傳分析工具(RETRAN)當作輔佐工具，來輔佐此TRACE模型，本論文中使用雷傳分析工具輔助進行此三座電廠之預期暫態未急停事故分析，並且擷取雷傳計算結果當作TRACE模型的輸入檔；在此，我們驗證此TRACE模型所得到的結果將符合電廠終期安全分析報告FSAR對於抑壓池溫度的最大限值內。
此外，在事故探討上，本論文將討論抑壓池及替代性插棒系統、自動注硼系統等安全裝置對於核電廠安全的影響，此研判也可作為未來核一、核二、核四廠ATWS事故分析的參考。

ABSTRACT
In this thesis, we use the 1st, 2nd and 4th Nuclear Power Plants as our research studies. We use the TRACE tool to build a CONTAN module. We use the CONTAN module to simulate the suppression pool. When MSIV Closure occurs at 100 % power, the SRV’s will open and steam will flow into the suppression pool. This thesis focuses on the analysis of suppression pool with MSIV Closure occurring at 100 % power (ATWS).
Anticipated Transient Without Scram, abbreviated as ATWS, means an anticipated operational occurrence followed by failure of the Reactor Protection System. Anticipated operational occurrences (AOOs), are expected to occur one or more times during the life of a nuclear power unit. The integrated ATWS prevention and mitigation features are designated as a Redundant Reactivity Control System. In this thesis, we will focus on (1) Alternate Rod Insertion (ARI) and Fine Motion Control Drive (FMCRD), and (2) automatic SLCS.
By using the TRACE CONTAN module, we will prove that the test results of ATWS performance evaluation stay within FSAR and design margin. We will use RETRAN as an auxiliary tool to verify our studies. We will use the RETRAN output of steam flow from SRV with MSIV Closure occurring at 100 % power ATWS as the input for the TRACE model. We will verify that our results stay within the most limiting values.
In conclusion, we will analyze the impacts of ARI, FMCRD, and SLCS on nuclear power plant safety. This thesis can also be used as a reference for Chinshan, Kuosheng and Lungmen plants’ ATWS analysis in the future.

TABLE OF CONTENTS
摘 要    I
ABSTRACT    II
致 謝ACKNOWLEDGEMENT    III
ACRONYMS AND ABBREVIATION    IV
TABLE OF CONTENTS    V
LIST OF TABLES    VII
LIST OF FIGURES    VII
1. INTRODUCTION    1
2. DESCRIPTION OF TRACE CODE    2
2.1 OVERVIEW    2
2.2 INSTALLATION AND CONFIGURATION    4
2.2.1 STEP 1 - Install jEdit and the SNAP-jEdit plug-in (optional)    4
2.2.2 STEP 2 - Install the APT Plotting Software (optional)    6
2.3 THEOREM OF RELAVENT COMPONENTS    18
2.3.1 TRACE COMPONENT “FILL”    18
2.3.2 TRACE COMPONENT “BREAKER”    24
2.3.3 TRACE COMPONENT “PIPE”    34
2.3.4 TRACE COMPONENT “CONTAN”    38
3. SIMULATION METHODS    63
3.1 TRACE BUILDING MODEL    67
3.2 INPUT DESCRIPTIONS    71
4. RESULTS AND DISCUSSIONS    74
4.1 TRANSIENT SCENARIOS DESCRIPTIONS    74
4.1.1 CHINSHAN  NUCLEAR POWER PLANT    75
4.1.1.1 MITIGATION METHOD : ACTIVATION OF SLCS    78
4.1.2 KUOSHENG NUCLEAR POWER PLANT    84
4.1.2.1 MITIGATION METHOD : ACTIVATION OF SLCS    88
4.1.3 LUNGMEN NUCLEAR POWER PLANT    94
4.1.3.1 MITIGATION METHOD : ACTIVATION OF ARI    95
4.1.3.2 MITIGATION METHOD : ACTIVATION OF FMCRD    98
4.1.3.3 MITIGATION METHOD : ACTIVATION OF SLCS    101
4.2 SUPPRESSION POOL TEMPERATURE TRANSIENT    104
4.2.1 CHINSHAN NUCLEAR POWER PLANT    104
4.2.1.1 MITIGATION METHOD: ACTIVATION OF  SLCS    105
4.2.2 KOUSHENG NUCLEAR POWER PLANT    110
4.2.2.1 MITIGATION METHOD: ACTIVATION OF  SLCS    110
4.2.3 LUNGMEN NUCLEAR POWER PLANT    116
4.2.3.1 MITIGATION METHOD: ACTIVATION OF  ARI    116
4.2.3.2 MITIGATION METHOD: ACTIVATION OF  FMCRD    122
4.2.3.3 MITIGATION METHOD: ACTIVATION OF  SLCS    127
4.3 RESULTS COMPARISON TABLES    138
5. CONCLUSIONS    140
REFERENCE    142
APPENDIX A –TRACE CODE FOR SIMULATION STUDY    144
LIST OF TABLES

Table 2-1 Example Input for CONTAN Component    43
Table 2-2 Volume versus Depth Table for Compartment 51 in Table 2-1    47
Table 2-3 Volume versus Depth Table for Compartment 52 in Table 2-1    48
Table 2-4 Cooler Temperature Table for Example Given in Table 2-1    54
Table 2-5 Example Input for BREAK Component Type 7.    61
Table 2-6 Example Input for FILL Component Type 7.    62
Table 4-2 Lungmen MSIV Closure Summary (FMCRD Run-In, EOC)    134
Table 4-3 Lungmen MSIV Closure Summary (Boron Injection, EOC)    134
Table 4-4 Thermal Power per Suppression Pool Volume    138
Table 4-5 Chinshan Suppression Pool Temperature Comparison    138
Table 4-6 Kousheng Suppression Pool Temperature Comparison    138
Table 4-7 Lungmen Suppression Pool Temperature Comparison (ARI)    139
Table 4-8 Lungmen Suppression Pool Temperature Comparison (FMCRD)    139
Table 4-9 Lungmen Suppression Pool Temperature Comparison (SLCS)    139

LIST OF FIGURES

Figure 1 SNAP TRACE MODEL EDIT DISPLAY    3
Figure 2  jEdit DISPLAY    5
Figure 3 APT Plotting Tools DISPLAY    8
Figure 4 SNAP TRACE CONFIGURATION DISPLAY    9
Figure 5 SNAP TRACE JOB STATUS DISPLAY    11
Figure 6 SNAP TRACE JOB SUMITTED DISPLAY    15
Figure 7 SNAP TRACE JOB SUMITTED CONSOLE DISPLAY    16
Figure 8 BREAK-component nodding diagram    21
Figure 9 FILL-component nodding diagram    23
Figure 10 TRACE Unchoked Flow Assessment Model    28
Figure 11  TRACE Choked Flow Assessment Mode    33
Figure 12  PIPE-component nodding diagram    37
Figure 13  A Typical BWR Containment    39
Figure 14  Compartment Spilling Model    49
Figure 15  The size of Lungmen Suppression Pool    65
Figure 16  Lungmen Main Steam System    66
Figure 17  TRACE Over All Diagram    68
Figure 18  The SRV on Steam Line    68
Figure 19  TRACE Suppression Pool Model    69
Figure 20  TRACE Component FILL Coupled With CONTAN    70
Figure 21  TRACE Components    73
Figure 22 Schematics of Chinshan Plant RETRAN System Thermal-Hydraulic Model    81
Figure 23 Schematics of Chinshan Plant RETRAN System Thermal-Hydraulic Model SRV Steam Flow    82
Figure 24 Schematics of Chinshan Plant RETRAN System Thermal-Hydraulic Model SRV Pressure    83
Figure 25 Schematics of Kuosheng Plant RETRAN System Thermal-Hydraulic Model    91
Figure 26 Schematics of Kuosheng Plant RETRAN System Thermal-Hydraulic Model SRV Steam Flow    92
Figure 27 Schematics of Kuosheng Plant RETRAN System Thermal-Hydraulic Model SRV Pressure    93
Figure 28  The Steam Flow from SRV after ARI    96
Figure 29  The Pressure of SRV after ARI    97
Figure 30  The Steam Flow from SRV after FMCRD    99
Figure 31  The Pressure of SRV after FMCRD    100
Figure 32  The Steam Flow from SRV after SLCS    102
Figure 33  The Pressure of SRV after SLCS    103
Figure 34  Chinshan Suppression Pool Temperature    106
Figure 35  Chinshan SHEX-05A Suppression Pool Temperature    108
Figure 36  Kuosheng Suppression Pool Temperature    112
Figure 37   Kuosheng SHEX-05A Suppression Pool Temperature    114
Figure 38  The Suppression pool temperature of ARI    118
Figure 39  Lungmen STEMP Suppression Pool Temperature (ARI, EOC)    120
Figure 40 The Suppression pool temperature of FMCRD Run-In    123
Figure 41  Lungmen STEMP Suppression Pool Temperature (FMCRD, EOC)    125
Figure 42  The Suppression pool temperature of SLCS injection    129
Figure 43  ARI, FMCRD and SLCS    130
Figure 44  Lungmen STEMP Suppression Pool Temperature (SLCS, EOC)    132
Figure 45  Lungmen MSIV Closure Summary (ARI, EOC)    135
Figure 46 Lungmen MSIV Closure Summary (FMCRD Run-In, EOC)    136
Figure 47 Lungmen MSIV Closure Summary (Boron Injection, EOC)    137

                                

[1] “TRACE V4.160 USER’S MANUAL VOLUME 2”, 2007, Division of Systems Analysis and Regulatory Effectiveness Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, DC 20555-0001.
[2] “Lungmen Unit 1 & 2 Final Safety Analysis Report Draft”, Aug. 2006, Taiwan Power Company.
[3] “Chinshan Plant Units 1 & 2 Final Safety Analysis Report”, sections 7.6.2.15 (RRCS, December 1993, Amendment 6) and 15.1.34 (Anticipated Transient Without Scram, December 1993, Amendment 6), Taiwan Power Company.
[4] “Evaluation of Chinshan ATWS Performance Final report”, GENE-770-39-0991, September 1991, General Electric Company.
[5] “NEDC-32039P”, revision 1, Class III, DRF A00-05010, maximum Extended load Line Limit and ARTS Improvement Program Analyses for Chinshan Nuclear Power Station Unit 1 and 2, November 1992, General Electric Company.
[6] “Kousheng Plant Units 1 & 2 Final Safety Analysis Report”, sections
7.6.1.14.2 (RRCS, December 2001, Amendment 12) and 15.8 (Anticipated Transient Without Scram, October 2004, Amendment 14), Taiwan Power Company.
[7] Kuosheng Nuclear Power Station Performance of ATWS Mitigation System
at Rated Power Operation and at Maximum Extended Operating Domain,
GENE-770-16-0692, June 1992, General Electric Company.
[8] 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants.
[9] 10CFR50.62, Requirements for Reduction of Risk from Anticipated
Transient without Scram (ATWS) Events for Light Water Cooled Nuclear
Power Plants, June 24, 1984.

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