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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 저널정보
- 대한기계학회 Journal of Mechanical Science and Technology Journal of Mechanical Science and Technology Vol.21 No.4
- 발행연도
- 2007.4
- 수록면
- 654 - 663 (10page)
이용수
초록· 키워드
오류제보하기
Nuclear power plant components suffer pipe shedding, cracking, thermal fatigue, bending and supporting bracket breakage during their life span. Notably, the horizontal inlet nozzle of steam generator is prone to thermal stratification frequently due to its operational characteristics. As a result, PWRs in many countries including the U.S.A. suffered a lot of pipe cracks and leakages around the late 1970s, as the thermal stress inflicted by thermal stratification formed in the horizontal inlet nozzle of steam generator during transition (auxiliary feedwater injection) was not reflected on power plant design.
Therefore, we classified the nuclear power plants in Korea into KSNP and Westinghouse-type(W) power plants (Kori # 1,2,3,4, Yeonggwang # 1,2 and Uljin # 1,2) and conducted an experiment on thermal stratification and thermal cycling in relation to the volume of auxiliary feedwater injected into the horizontal inlet nozzle of steam generator and hot water flowing back from steam generator. As a result, it was found out that KSNP was hardly prone to thermal stratification while thermal stratification occurred in the horizontal inlet nozzle of steam generator in Westinghousetype(W) power plants, necessitating a solution to address such a phenomenon.
Therefore, we classified the nuclear power plants in Korea into KSNP and Westinghouse-type(W) power plants (Kori # 1,2,3,4, Yeonggwang # 1,2 and Uljin # 1,2) and conducted an experiment on thermal stratification and thermal cycling in relation to the volume of auxiliary feedwater injected into the horizontal inlet nozzle of steam generator and hot water flowing back from steam generator. As a result, it was found out that KSNP was hardly prone to thermal stratification while thermal stratification occurred in the horizontal inlet nozzle of steam generator in Westinghousetype(W) power plants, necessitating a solution to address such a phenomenon.
목차
Abstract
1. Introduction
2. T/S cases & operational conditions
3. Test rig & test method
4. Experiment results & discussions
5. Conclusions
6. Recommendations
Acknowledgements
References
참고문헌 (15)
참고문헌 신청
Design Report No. Cenc-1940 Revision 01(Z-160-RA-C01-01), 1986, “Design Report For YONG-GWANG Nuclear Power Plant 3 And 4 Stream Generators.”
EPRI, 1994, "Thermal Stratification, Cycling and Striping(TASCS),” TR-103581.
KINS, KINS/Ar-267, 1994, "Numerical Analysis of Thermal Stratification Inside the Horizontal Pipe(Ι).”
P&ID Reactor Coolant System, 9-431-N105-001006.
P&ID Safety Injection/Shutdown Cooling System, 9-441-N105-001-004.
EPRI, 1994, "Thermal Stratification, Cycling and Striping(TASCS),” TR-103581.
KINS, KINS/Ar-267, 1994, "Numerical Analysis of Thermal Stratification Inside the Horizontal Pipe(Ι).”
P&ID Reactor Coolant System, 9-431-N105-001006.
P&ID Safety Injection/Shutdown Cooling System, 9-441-N105-001-004.
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