核嚴重事故分析程式研究及其應用

Study on Nuclear Severe Accident Codes and Their Applications

王德全(Wang-Te Chuan)

指導教授 ： 鄧治東

中原大學/工學院/機械工程學系/博士(2007年)

https://doi.org/10.6840/cycu200700481

摘要

本篇論文主要研究核嚴重事故分析程式及其應用，內容主要分為四部份:
1. 嚴重事故程式的比較:
使用SCDAP/RELAP5(SR5)、MAAP 及MELCOR程式對核二、三廠進行大破口冷卻水流失事故(LOCA)及全黑事故(SBO)序列進行分析與比對。在核二廠大破口冷卻水流失事故MAAP分析出較長之爐底失效時間，而MELCOR預測出較短之爐底失效時間。在核三廠全黑事故事故，與SR5及MELCOR程式比較，MAAP分析出較長之蒸汽產生器(SG)蒸乾時間、水位由實際燃料頂部(TAF)到實際燃料底部(BAF)時距及爐底失效時間。
2.嚴重事故處理引的驗證
根據一廠安全度評估(PRA)報告，以爐心熔毀最高機率的事故(T5UtXC)為驗證的案例。根據MAAP程式分析結果，在T5UtXC事故RPV水位下降到1/4爐心水位前採取緊急洩壓，可維持爐心適當的冷卻，兩個控制棒泵流量即可維持RPV水位高於實際燃料頂端。
3.生水系統救援能力的探討
核一廠安全度評估並未考量山上生水系統的救援能力，根據MAAP程式的分析結果，在T3UTERDG事故生水系統無法冷卻爐心，在反應器壓力槽失效後，在T3UTERDGX事故生水系統無法將乾井水位灌到最低熔渣淹蓋水位(MDSL)之上，靈敏度分析顯示，在T3UTERDG事故，RPV水位下降到二階水位(L-2)前，起動生水系統可維持爐心適當的冷卻，在T3UTERDGX事故，三倍生水系統注水流量為將乾井水位灌到最低熔渣淹蓋水位及冷卻乾井地板熔渣的最低流量。
4. 反應爐水位指示系統的研究:
核三廠在2001年3月18日發生了電廠全黑事故，以MELCOR程式模擬核三廠全黑事故。另外針對RVLIS幾個有趣的現象進行研究，包括起始RPV滿水位超過100%、壓力槽水位收縮(shrinkage)及Upper Range A串水位計有兩個突升(peak)，起始滿水位112%主要因為水位儀器的校正，全黑事故後1.2小時水位下降，主要因為爐水達到飽和，全黑事故後Upper Range A串水位計有兩個突升，是因為調壓槽的飽和水流到第二迴路的熱管，結果顯示MELCOR程式可模擬出RVLIS的反應。

關鍵字

嚴重事故 ； 沸水式反應器 ； 嚴重事故處理指引 ； 壓水式反應器

並列摘要

This thesis demonstrates the study of severe accident codes and their applications. It divides into four parts.
1. Severe Accident Codes Comparison:
Large break loss off coolant accident (LOCA) sequence of Kuosheng nuclear power plant (NPP) and station blackout (SBO) sequence of Maanshan NPP with SCDAP/RELAP5 (SR5), MAAP and MELCOR codes are studied. MAAP predicts a longer vessel failure time and MELCOR predicts a shorter vessel failure time for the large break LOCA sequence. MAAP predicts a longer time of steam generator (SG) dryout, time interval between top of active fuel (TAF) and bottom of active fuel (BAF) and vessel failure time than those of SR5 and MELCOR prediction for the station blackout sequence.
2. Severe Accident Management Guidelines Validation
T5UtXC sequence, the highest core melt frequency (CDF) in probabilistic risk assessment (PRA) insight of Chinshan NPP, is cited as a reference case for SAMGs validation. Based on MAAP4 calculation, the result shows that RPV depressurization before the reactor water level reaches one fourth of the core water level can prevent the core from damage in the T5UtXC sequence. Flow rate of two control rod drive (CRD) pumps is enough to maintain the reactor water level above TAF and cool down the core in the T5UtXC sequence without operator action.
3. The Raw Water System Performance
No credit is taken for raw water system in the development of Chinshan PRA. Based on MAAP4 analysis, the raw water system can’t cool down the core in the T3UTERDG sequence after entering SAMGs. The raw water system can’t flood drywell water level above minimum debris submerge level (MDSL) in the T3UTERDGX sequence after reactor pressure vessel (RPV) breach. The sensitivity studies show raw water injection before the vessel water level reached Level 2 (L-2) can keep core coolability in the T3UTERDG sequence. Three times of raw water injection rate is the minimum flow rate to flood the drywell water level above MDSL and cool down the corium on the drywell floor in the T3UTERDGX sequence.
4. RVLIS Investigation
A station blackout incident occurred in the Maanshan NPP on March 18, 2001. The station blockout incident was simulated with MELCOR 1.8.5 code. The interesting phenomena about RVLIS responses included initial full water level above 100%, reactor pressure vessel water shrinkage, and two peaks in upper range train A. The initial full water levels of upper range were at about 112% because of calibration conditions. The two trains of upper range dropped at about 1.2 hr after SBO because of RPV water saturation. RVLIS upper range train A had two level rises after SBO because of two flows out of pressurizer (PZR) into loop 2 hot leg.

並列關鍵字

Severe Accident Management Guidelines ； BWR ； PWR ； Severe Accident

參考文獻

17.Wang TC, Wang SJ Wang and Teng JT, “Analysis of The Chinshan Raw Water System Performance for Severe Accident” Nucl. Tech, 156 (3) 347-359 DEC 2006.

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8.“Final Safety Analysis Report of the Maanshan Nuclear Power Station,” Taiwan Power Company (1982).