A computational fluid dynamics (CFD) analysis was performed to investigate the hydraulic response of the flow field inside the pressurized water reactor (PWR) steam generator (SG) secondary side and the connected part of main feed water pipe to an abrupt main feed water line break (FWLB) accident. To realistically analyze the transient flow field situation, the flow field was assumed to be occupied initially by highly compressed subcooled water except that the upper part of the SG secondary side where steam occupied as in the practical case and the break was assumed to occur at the circumferential weld line between the feed water nozzle and the main feed water pipe. This would result in a subcooled water flashing flow from the SG through the short-broken pipe end to the surrounding atmosphere, which was numerically simulated in this study. Typical results of the prediction in terms of the fluid transient velocity and pressure were illustrated and discussed. To examine the physical validity of the present numerical simulation of the subcooled water flashing flow, the transient mass flow rates predicted in this study were compared with the other previous numerical predictions based on the subcooled water nonflashing (no phase change) flow or saturated water flashing flow assumptions and the prediction by a simple analysis method.
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August 2019
Technical Briefs
Numerical Analysis of Subcooled Water Flashing Flow From a Pressurized Water Reactor Steam Generator Through an Abruptly Broken Main Feed Water Pipe
Jong Chull Jo,
Jong Chull Jo
School of Mechanical Engineering,
Pusan National University,
63 Busandaehak-ro,
Busan, Geumjeong-gu 46241, South Korea;
Pusan National University,
63 Busandaehak-ro,
Busan, Geumjeong-gu 46241, South Korea;
Reactor System Evaluation Department,
Korea Institute of Nuclear Safety,
62 Gwahak-ro,
Daejon, Yusung-gu 34142, South Korea
e-mail: jongcjo@pusan.ac.kr
Korea Institute of Nuclear Safety,
62 Gwahak-ro,
Daejon, Yusung-gu 34142, South Korea
e-mail: jongcjo@pusan.ac.kr
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Jae Jun Jeong,
Jae Jun Jeong
School of Mechanical Engineering,
Pusan National University,
Busan,
e-mail: jjjeong@pusan.ac.kr
Pusan National University,
63 Busandaehak-ro
,Busan,
Geumjeong-gu
46241, South Koreae-mail: jjjeong@pusan.ac.kr
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Byong Jo Yun,
Byong Jo Yun
School of Mechanical Engineering,
Pusan National University,
Busan,
e-mail: bjyun@pusan.ac.kr
Pusan National University,
63 Busandaehak-ro
,Busan,
Geumjeong-gu
46241, South Koreae-mail: bjyun@pusan.ac.kr
Search for other works by this author on:
Jongkap Kim
Jongkap Kim
Reactor System Evaluation Department,
Korea Institute of Nuclear Safety,
e-mail: k233kjk@kins.re.kr
Korea Institute of Nuclear Safety,
62 Gwahak-ro
, Daejon,Yusung-gu
34142, South Koreae-mail: k233kjk@kins.re.kr
Search for other works by this author on:
Jong Chull Jo
School of Mechanical Engineering,
Pusan National University,
63 Busandaehak-ro,
Busan, Geumjeong-gu 46241, South Korea;
Pusan National University,
63 Busandaehak-ro,
Busan, Geumjeong-gu 46241, South Korea;
Reactor System Evaluation Department,
Korea Institute of Nuclear Safety,
62 Gwahak-ro,
Daejon, Yusung-gu 34142, South Korea
e-mail: jongcjo@pusan.ac.kr
Korea Institute of Nuclear Safety,
62 Gwahak-ro,
Daejon, Yusung-gu 34142, South Korea
e-mail: jongcjo@pusan.ac.kr
Jae Jun Jeong
School of Mechanical Engineering,
Pusan National University,
Busan,
e-mail: jjjeong@pusan.ac.kr
Pusan National University,
63 Busandaehak-ro
,Busan,
Geumjeong-gu
46241, South Koreae-mail: jjjeong@pusan.ac.kr
Byong Jo Yun
School of Mechanical Engineering,
Pusan National University,
Busan,
e-mail: bjyun@pusan.ac.kr
Pusan National University,
63 Busandaehak-ro
,Busan,
Geumjeong-gu
46241, South Koreae-mail: bjyun@pusan.ac.kr
Jongkap Kim
Reactor System Evaluation Department,
Korea Institute of Nuclear Safety,
e-mail: k233kjk@kins.re.kr
Korea Institute of Nuclear Safety,
62 Gwahak-ro
, Daejon,Yusung-gu
34142, South Koreae-mail: k233kjk@kins.re.kr
Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received August 17, 2018; final manuscript received March 17, 2019; published online April 25, 2019. Assoc. Editor: Victor P. Janzen.
J. Pressure Vessel Technol. Aug 2019, 141(4): 044501 (10 pages)
Published Online: April 25, 2019
Article history
Received:
August 17, 2018
Revised:
March 17, 2019
Citation
Jo, J. C., Jeong, J. J., Yun, B. J., and Kim, J. (April 25, 2019). "Numerical Analysis of Subcooled Water Flashing Flow From a Pressurized Water Reactor Steam Generator Through an Abruptly Broken Main Feed Water Pipe." ASME. J. Pressure Vessel Technol. August 2019; 141(4): 044501. https://doi.org/10.1115/1.4043297
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