TAILIEUCHUNG - Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. | Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes EPJ Nuclear Sci. Technol. 2 16 2016 Nuclear Sciences . Park and . Song published by EDP Sciences 2016 amp Technologies DOI epjn 2016010 Available online at http REGULAR ARTICLE Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes Joo Hwan Park and Yong Mann Song Korea Atomic Energy Research Institute 989-111 Daedukdaero Yuseong-gu Taejon 305-353 Korea Received 30 September 2015 Received in final form 20 January 2016 Accepted 4 February 2016 Published online 1 April 2016 Abstract. A CANDU-6 reactor which has 380 fuel channels of a pressure tube type is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a

• Năng lượng
• Crept pressure tubes
• Inner pitch length modification
• 37 element fuel bundle
• CANDU 6 reactor
• Critical Heat Flux
• Tight rod bundle
• Wire spacer
• Pitch to diameter ratio
• Two phase flow
• báo cáo khoa học
• báo cáo hóa học
• công trình nghiên cứu về hóa học
• tài liệu về hóa học
• cách trình bày báo cáo
• Elevated temperatures
• Incident heat flux Thermal boundary condition
• Concrete deformation
• Revision of concrete performance