CANDU-9 원자로 열수송계통의 과도변화해석 Transient Analysis of the Heat Transport System in CANDU-9 Reactor
한국기계기술학회지
(韓國機械技術學會誌)
제12권 제3호
(2010.09)
pp.59-65
pp.59-65
한국기계기술학회
(Korean Society of Mechanical Technology)
초록
The Heavy Water Reactor(HWR) Heat Transport(HT) system transient analysis for the design of major nuclear equipment during normal and abnormal operating conditions was performed. The compliance with requirements of AECB Regulatory Document R-77 for CANDU reactor was estimated in CANDU-9 nuclear reactor. The analysis results showed that for each postulated accident the peak pressure values in the reactor headers are within the acceptance criteria given in ASME code requirements and the fuel overheating is prevented. The analysis results showed that the flow reversal through the fuel channel occurred but didn't result in any damage on the fuel bundle.
키워드
목차
Abstract
1. 서론
2. CANDU-9 열수송계통 형상
3. 설계요구사항
4. 코드 모델링
5. 결과 및 논의
5.1 정상운전상태
5.2 One Pump Start-up
5.3 Reactor Trip
5.4 Loss of Class IV Power
5.5 Pump Shaft Seizure
6. 결론
참고문헌
1. 서론
2. CANDU-9 열수송계통 형상
3. 설계요구사항
4. 코드 모델링
5. 결과 및 논의
5.1 정상운전상태
5.2 One Pump Start-up
5.3 Reactor Trip
5.4 Loss of Class IV Power
5.5 Pump Shaft Seizure
6. 결론
참고문헌
저자