scholarly journals Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

2012 ◽  
Vol 49 (8) ◽  
pp. 857-866 ◽  
Author(s):  
Youngsuk Bang ◽  
Gunhyo Jung ◽  
Byungchul Lee ◽  
Kwang-Il Ahn
2020 ◽  
Vol 57 (12) ◽  
pp. 1287-1296
Author(s):  
Naoya Miyahara ◽  
Shuhei Miwa ◽  
Mélany Gouëllo ◽  
Junpei Imoto ◽  
Naoki Horiguchi ◽  
...  

2012 ◽  
Vol 614-615 ◽  
pp. 626-631
Author(s):  
Chang Hong Peng ◽  
Ying Hao Yang

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg and SG tubes. The first step performs thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step applies the creep rupture models to test the potential for the degraded SG to rupture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in LERF (Large and Early Release Frequency) model, and the overall LERF risk due to the Induced SGTR was calculated. In the final step, implementation of severe accident management guidance (SAMG), such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the result of induced SGTR and LERF effectively.


1990 ◽  
Author(s):  
T.J. Heames ◽  
D.A. Williams ◽  
N.A. Johns ◽  
N.M. Chown ◽  
N.E. Bixler ◽  
...  

Author(s):  
Jongmin Kim ◽  
Min-Chul Kim ◽  
Joonyeop Kwon

Abstract The materials used previously for steam generator tubes around the world have been replaced and will be replaced by Alloy 690 given its improved corrosion resistance relative to that of Alloy 600. However, studies of the high- temperature creep and creep-rupture characteristics of steam generator tubes made of Alloy 690 are insufficient compared to those focusing on Alloy 600. In this study, several creep tests were conducted using half tube shape specimens of the Alloy 690 material at temperatures ranging from 650 to 850C and stresses in the range of 30 to 350 MPa, with failure times to creep rupture ranging from 3 to 870 hours. Based on the creep test results, creep life predictions were then made using the well-known Larson Miller Parameter method. Steam generator tube rupture tests were also conducted under the conditions of a constant temperature and pressure ramp using steam generator tube specimens. The rupture test equipment was designed and manufactured to simulate the transient state (rapid temperature and pressure changes) in the event of a severe accident condition. After the rupture test, the damage to the steam generator tubes was predicted using a creep rupture model and a flow stress model. A modified creep rupture model for Alloy 690 steam generator tube material is proposed based on the experimental results. A correction factor of 1.7 in the modified creep rupture model was derived for the Alloy 690 material. The predicted failure pressure was in good agreement with the experimental failure pressure.


2019 ◽  
Vol 348 ◽  
pp. 14-23 ◽  
Author(s):  
JinHo Song ◽  
ByungHee Lee ◽  
SungIl Kim ◽  
GwangSoon Ha

Author(s):  
Osamu Kawabata ◽  
Masao Ogino

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.


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