scholarly journals Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

2016 ◽  
Vol 48 (4) ◽  
pp. 952-964 ◽  
Author(s):  
Sang June Ahn ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Seok Hun Kang ◽  
Kwi Lim Lee ◽  
...  
Keyword(s):  
Steam Generator ◽  
Fast Reactor ◽  
Reaction Event ◽  
Steam Generator Tubes

Parametric Study of External Pressure on Steam Generator Tube Bends

2012 ◽  
Author(s):  
Mitch Hokazono ◽  
Clayton T. Smith
Keyword(s):  
Steam Generator ◽  
Parametric Study ◽  
External Pressure ◽  
Light Water Reactor ◽  
Material Strength ◽  
Steam Generators ◽  
Elliptical Cross ◽  
Light Water ◽  
Water Reactor ◽  
Steam Generator Tubes

Integral light-water reactor designs propose the use of steam generators located within the reactor vessel. Steam generator tubes in these designs must withstand external pressure loadings to prevent buckling, which is affected by material strength, fabrication techniques, chemical environment and tube geometry. Experience with fired tube boilers has shown that buckling in boiler tubes is greatly alleviated by controlling ovality in bends when the tubes are fabricated. Light water reactor steam generator pressures will not cause a buckling problem in steam generators with reasonable fabrication limits on tube ovality and wall thinning. Utilizing existing Code rules, there is a significant design margin, even for the maximum differential pressure case. With reasonable bend design and fabrication limits the helical steam generator thermodynamic advantages can be realized without a buckling concern. This paper describes a theoretical methodology for determining allowable external pressure for steam generator tubes subject to tube ovality based on ASME Section III Code Case N-759-2 rules. A parametric study of the results of this methodology applied to an elliptical cross section with varying wall thicknesses, tube diameters, and ovality values is also presented.

Evaluation of Motorized Rotating Pancake Coil Probe Signals Simulated by Numerical Analysis in Steam Generator Tubes

2006 ◽  
Vol 321-323 ◽  
pp. 451-454
Author(s):  
Joo Young Yoo ◽  
Sung Jin Song ◽  
Chang Hwan Kim ◽  
Hee Jun Jung ◽  
Young Hwan Choi ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Numerical Analysis ◽  
Steam Generator ◽  
Integral Method ◽  
Volume Integral ◽  
Analysis Software ◽  
Steam Generator Tubes ◽  
Calibration Tool ◽  
High Possibility ◽  
Good Agreement

In the present study, the synthetic signals from the combo tube are simulated by using commercial electromagnetic numerical analysis software which has been developed based on a volume integral method. A comparison of the simulated signals to the experiments is made for the verification of accuracy, and then evaluation of five deliberated single circumferential indication signals is performed to explore a possibility of using a numerical simulation as a practical calibration tool. The good agreement between the evaluation results for two cases (calibration done by experiments and calibration made by simulation) demonstrates such a high possibility.

Evaluation of the Creep Rupture Behavior of Alloy 690 Steam Generator Tubes Considering the Pressure Ramp Rate

2021 ◽  
Author(s):  
Jongmin Kim ◽  
Min-Chul Kim ◽  
Joonyeop Kwon
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Parameter Method ◽  
Alloy 600 ◽  
Steam Generator Tube ◽  
Failure Pressure ◽  
Alloy 690 ◽  
Rupture Model ◽  
Temperature And Pressure ◽  
Steam Generator Tubes

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.

A comparative study on the fretting wear of steam generator tubes in korean power plants

Wear ◽  
2003 ◽  
Vol 255 (7-12) ◽  
pp. 1198-1208 ◽  
Author(s):  
Young-Ho Lee ◽  
Hyung-Kyu Kim ◽  
Hong-Deok Kim ◽  
Chi-Yong Park ◽  
In-Sup Kim
Keyword(s):  
Comparative Study ◽  
Steam Generator ◽  
Power Plants ◽  
Fretting Wear ◽  
Steam Generator Tubes

scholarly journals A REVIEW ON THE ODSCC OF STEAM GENERATOR TUBES IN KOREAN NPPS

2013 ◽  
Vol 45 (4) ◽  
pp. 513-522 ◽  
Author(s):  
HANSUB CHUNG ◽  
HONG-DEOK KIM ◽  
SEUNGJIN OH ◽  
MYUNG HWAN BOO ◽  
KYUNG-HWAN NA ◽  
...  
Keyword(s):  
Steam Generator ◽  
Steam Generator Tubes
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