Review of the Effect of Hydraulic Load and Stress on the Feedwater Box Depending on the Location of Break during APR1400 Steam Generator Feedwater Line Break Accident

被引:0
|
作者
Jung, Han Sik [1 ]
Kim, Jong In [1 ]
机构
[1] Doosan Enerbil, Nucl Component Design Team, Changwon Si, South Korea
关键词
Steam Generator; Feedwater Line Break; Hydraulic Load; Stress; Location of Break;
D O I
10.3795/KSME-A.2023.47.12.1039
中图分类号
TH [机械、仪表工业];
学科分类号
0802 ;
摘要
The APR1400 steam generator requires a design to secure the integrity of the internal structure during a feedwater line break accident, which is a design-based accident, and an evaluation is conducted to confirm this. When the feedwater line break occurs, the fluid inside the steam generator is momentarily released and the pressure drops rapidly, resulting in a hydraulic load on the internal structure. The accident analysis code is used to calculate the hydraulic load, and the break accident requires the end of the feedwater line nozzle as the location of the accident. This study analyzes the effect of the hydraulic load and stress on the feedwater box depending on the location of the feedwater line break accident.
引用
收藏
页码:1039 / 1045
页数:7
相关论文
共 9 条
  • [1] THE STUDY OF THERMAL-HYDRAULIC ANALYSIS FOR FEEDWATER LINE BREAK IN STEAM GENERATOR
    Jang, Minji
    Kim, Jongin
    PROCEEDINGS OF THE 20TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING AND THE ASME 2012 POWER CONFERENCE - 2012, VOL 1, 2012, : 603 - 609
  • [2] Experimental study on the blowdown load during the steam generator feedwater line break accident in the evolutionary pressurized water reactor
    Kang, Kyoung-Ho
    Park, Hyun-Sik
    Cho, Seok
    Choi, Nam-Hyun
    Bae, Sung-Won
    Lee, Seung-Wook
    Kim, Yeon-Sik
    Choi, Ki-Yong
    Baek, Won-Pi
    Kim, Moo-Yong
    ANNALS OF NUCLEAR ENERGY, 2011, 38 (05) : 953 - 963
  • [3] Main steam line break accident simulation of APR1400 using the model of ATLAS facility
    Ekariansyah, A. S.
    Deswandri
    Sunaryo, Geni R.
    INTERNATIONAL CONFERENCE ON NUCLEAR ENERGY TECHNOLOGIES AND SCIENCES (ICONETS 2017), 2018, 962
  • [4] Numerical prediction of transient hydraulic loads acting on PWR steam generator tubes and supports during blowdown following a feedwater line break
    Jo, Jong Chull
    Jeong, Jae Jun
    Yun, Byong Jo
    Kim, Jongkap
    NUCLEAR ENGINEERING AND TECHNOLOGY, 2021, 53 (01) : 322 - 336
  • [5] INTEGRAL EFFECT TESTS ON TRANSIENT THERMAL-HYDRAULIC BEHAVIOR DURING A STEAM GENERATOR TUBE RUPTURE ACCIDENT IN THE APR1400
    Kang, Kyoung-Ho
    Park, Hyun-Sik
    Cho, Seok
    Choi, Nam-Hyun
    Chu, In-Cheol
    Yun, Byong-Jo
    Kim, Kyung-Doo
    Kim, Yeon-Sik
    Baek, Won-Pil
    Choi, Ki-Yong
    NUCLEAR TECHNOLOGY, 2012, 177 (03) : 382 - 394
  • [6] Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident
    Bae, Byoung-Uhn
    Kim, Seok
    Park, Yu-Sun
    Kang, Kyoung-Ho
    NUCLEAR ENGINEERING AND DESIGN, 2014, 275 : 249 - 263
  • [7] Transient Hydraulic Response of a Pressurized Water Reactor Steam Generator to a Feedwater Line Break Using the Nonflashing Liquid Flow Model
    Jo, Jong Chull
    Jeong, Jae Jun
    Moody, Frederick J.
    JOURNAL OF PRESSURE VESSEL TECHNOLOGY-TRANSACTIONS OF THE ASME, 2017, 139 (03):
  • [8] Numerical and analytical predictions of nuclear steam generator secondary side fl ow fi eld during blowdown due to a feedwater line break
    Jo, Jong Chull
    Jeong, Jae-Jun
    Moody, Frederick J.
    NUCLEAR ENGINEERING AND TECHNOLOGY, 2021, 53 (03) : 1029 - 1040
  • [9] EFFECTS OF INITIAL PRESSURE AND LENGTH OF A BROKEN PIPE ON THE TRANSIENT HYDRAULIC LOADS ACTING ON NUCLEAR STEAM GENERATOR TUBES AND SUPPORTS DURING BLOWDOWN FOLLOWING A SUDDEN FEEDWATER PIPE BREAK
    Jo, Jong Chull
    Jeong, Jae-Jun
    Yun, Byong-Jo
    PROCEEDINGS OF THE ASME PRESSURE VESSELS AND PIPING CONFERENCE, 2019, VOL 4, 2019,