反應器壓力槽之老化評估研究

Abstract

反應器壓力槽為核能電廠蒸汽供應系統中極為重要的壓力邊界組件，在電廠長期運轉後，反應器壓力槽會面臨各種老化問題，而影響電廠之安全；其中，輻射脆化為反應器壓力槽最主要之老化機制。目前國內電廠係依據美國核能管制委員會所制訂的規範，以定論式的分析方法，針對反應器壓力槽之輻射脆化問題，定期進行老化評估，期能確保其在運轉期間之安全。本研究根基於現行定論式之評估模式，建構一套反應器壓力槽之機率式評估模式，據以估算壓力槽因輻射脆化而超出法規規定之可能性。本研究以國內壓水式核能電廠之反應器壓力槽為案例進行此項分析與評估，並針對其結果加以討論。研究結果顯示，在合理的假設下，若以「參考溫度」評估，該反應器壓力槽於設計壽命終期時，有3.42×10^7因輻射脆化至超出法規規定之可能性；但若以「上限衝擊能量」評估，則有1×10^4 超出法規規定之可能性。因為現行法規主要仍根基於傳統保守式的定論分析觀念，所以反應器壓力槽因輻射脆化而導致破壞的機率應仍小於以上數據。

The reactor pressure vessel (RPV) is an extremely important component in the steam supply system of a nuclear power plant. However, after a long period of operation, it will face various aging problems that would affect the safety of the plant. Among them, radiation embrittlement is considered the major aging mechanism of the reactor pressure vessels. To evaluate the degree of radiation embrittlement and ensure the safety of the vessels, the domestic nuclear power plants adopt the standard set by the United States Nuclear Regulatory Commission (USNRC) and carry out aging assessment programs periodically. Based on the deterministic assessment procedure as required by USNRC, the present study proposes a probabilistic aging assessment model to estimate the probability of outrunning the required values set by USNRC in consideration of radiation embrittlement of reactor pressure vessels. Taking the domestic pressurized water nuclear power plant as an example, the present study shows that, under a reasonable assumption and based on the consideration of adjusted reference temperature (ART), the reactor pressure vessel might outrunning the code-required value with a probability of 3.42×10^7 at the end of its design life. If the calculation is based on the consideration of upper shelf energy (USE), the probability would become 1×10^4 . Since the required values set by USNRC are based on the rather conservative deterministic consideration, the real failure probability of the studied vessel owing to radiation embrittlement must be lower than the above mentioned values.