@article{ART001486149},
author={ 신정철 },
title={CANDU-9 원자로 열수송계통의 과도변화해석},
journal={한국기계기술학회지},
issn={1229-604X},
year={2010},
volume={12},
number={3},
pages={59-65},
doi={10.17958/ksmt.12.3.201009.59},
url={http://dx.doi.org/10.17958/ksmt.12.3.201009.59}

TY - JOUR
AU - 신정철
TI - CANDU-9 원자로 열수송계통의 과도변화해석
T2 - 한국기계기술학회지
PY - 2010
VL - 12
IS - 3
PB - 한국기계기술학회
SP - 59-65
SN - 1229-604X
AB - 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.
KW - Reactor Operating Condition(원자로운전상태), ASME Code(미국기계학회법규), CANDU Reactor(캔두형 원자로), Heat Transport System(열수송계통)
DO - 10.17958/ksmt.12.3.201009.59
UR - http://dx.doi.org/10.17958/ksmt.12.3.201009.59
ER -

신정철 . 2010, “CANDU-9 원자로 열수송계통의 과도변화해석”, 한국기계기술학회지, vol. 12, no. 3, pp. 59-65. Available from: doi:10.17958/ksmt.12.3.201009.59

신정철 “CANDU-9 원자로 열수송계통의 과도변화해석” 한국기계기술학회지 12.3 pp. 59-65 (2010): 59.

신정철 . CANDU-9 원자로 열수송계통의 과도변화해석 한국기계기술학회지 [Internet]. 2010; 12( 3), : 59-65. Available from: doi:10.17958/ksmt.12.3.201009.59

신정철 . “CANDU-9 원자로 열수송계통의 과도변화해석” 한국기계기술학회지 12, no.3 (2010): 59-65. doi: 10.17958/ksmt.12.3.201009.59