Thermal Aging of Primary Circuit Piping Materials in PWR Nuclear Power Plant
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1215-V18-07
Thermal Aging of Primary Circuit Piping Materials in PWR Nuclear Power Plant Xitao Wang 1), Shilei Li 1), Shuxiao Li 1), Yanli Wang 1) Fei Xue 2), Guogang Shu 2) 1
State Key Laboratory for Advanced Metals and Materials, University of Science and Technology Beijing, 30 Xueyuan Rd. , Beijing 100083, China 2 China Guangdong Nuclear Power Group, Shenzhen 518028, China ABSTRACT The reserved cast austenitic stainless steels (CASS) for primary circuit piping in Daya Bay Nuclear Power Plant were studied. The changes of microstructure, mechanical properties and fracture behavior were investigated using SEM, EPMA, TEM and nanoindentation after accelerated aging at 400°C for up to 10000 h. Microhardness of ferrite increased rapidly in the early stage and then increased slowly later. The impact energy of materials declined with the aging time and reduced to a very low level after aging for 10000 hours. Fracture morphology displayed a mixture of cleavage in ferrite along with dimple and tearing in austenite. Two kinds of precipitations were observed in ferrite by TEM after long periods of aging. The fine Crenriched α′ phases precipitated homogeneously in ferrite, and a few larger G phases were observed as well. The precipitation of α′ phases was considered to be the primary mechanism of thermal aging embrittlement in CASS. INTRODUCTION
CASS materials are widely used in major components of PWR nuclear power plants due to their excellent strength, corrosion resistance and good weldability. The microstructure of CASS contains about 10-20% island ferrite in an austenite matrix. When they are used in piping system of primary cooling circuits, the service temperature is at the intermediate range of 280°C ~320°C, which is under its ductile-brittle transition temperature. However, when they are used in this temperature range for extended periods of time, they can suffer a loss of toughness [1]. Because no changes were observed in the austenite phases after long term aging, the embrittlement was considered to be associated with the changes in the ferrite phases [2]. Microstructure changes in the ferrite phases, for this type of aged cast stainless steels, were characterize by Chung and Chopra[3,4], Yamada[5], Auger[6], et al. Microstructural studies of these materials reported that phase decomposition to be the principal mechanism, wherein the ferrite decomposes into Fe-rich α and Cr-rich α ′ . G-phase precipitation in the ferrite was reported as secondary reactions [7-10]. Because realistic aging of component for end-of-life conditions at service temperature could not be produced, it was customary to simulate the metallurgical structure of a reactor component by accelerated aging at 400°C. Relative studies[11,12] indicated that the mechanisms of aging embrittlement were identical for the accelerated aging and reactor operating conditions.
EXPERIMENTAL DETAILS The material studied in this work is centrifugally CASS, containing 12% ferrite phase, which was cut from the primary coolant water pipe in Daya Bay Nuclear Power
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