CAJ | 학술논문

本文系统介绍了VVER‐1000型反应堆压力容器（RPV ）的温度监督情况，针对田湾核电站1＃机组RPV的温度监督测试结果进行分析，评价运行3年后RPV力学性能（包括拉伸、冲击、断裂韧性）变化行为及热老化脆化机理，评估了当前田湾RPV服役运行后的热老化脆化状态和温度监督的时间安排。结果表明，温度监督样品经过堆内高温环境考验后，焊缝材料表现出一定程度的脆化特征，但母材、热影响区脆化不明显。与康采恩模型的结果和俄罗斯数据相比较后，认为田湾核电站1＃机组RPV热老化脆化情况在合理范围内。
본문계통개소료VVER‐1000형반응퇴압력용기（RPV ）적온도감독정황，침대전만핵전참1＃궤조RPV적온도감독측시결과진행분석，평개운행3년후RPV역학성능（포괄랍신、충격、단렬인성）변화행위급열노화취화궤리，평고료당전전만RPV복역운행후적열노화취화상태화온도감독적시간안배。결과표명，온도감독양품경과퇴내고온배경고험후，한봉재료표현출일정정도적취화특정，단모재、열영향구취화불명현。여강채은모형적결과화아라사수거상비교후，인위전만핵전참1＃궤조RPV열노화취화정황재합리범위내。
VVER‐1000 reactor pressure vessel temperature surveillance is introduced systematically .The mechanical behaviors and thermal aging embrittlement mechanism of Tianwan unit 1 RPV after 3 years operation were analyzed and evaluated according to the results of temperature surveillance . The time scheme of next temperature surveillance was determined .The results show that after high temperature testing in reactor ,the sample weld for the temperature surveillance exhibits a certain degree of embrittlement ,but the embrittlement in the base metal and heat‐affected zone is not obvious .Comparing with Konzern model predicted value and the data from Russia ,it is considered that thermal aging embrittlement of RPV from Tianwan unit 1 is within a reasonable range .