Probabilistic assessment of excessive leakage through steam generator tubes degraded by secondary side corrosion

1998 ◽  
Vol 185 (2-3) ◽  
pp. 347-359 ◽  
Author(s):  
L. Cizelj ◽  
I. Hauer ◽  
G. Roussel ◽  
C. Cuvelliez
Keyword(s):  
Steam Generator ◽  
Probabilistic Assessment ◽  
Steam Generator Tubes ◽  
Secondary Side

CANTIA Code for the Probabilistic Assessment of Inspection Strategies for Steam Generator Tubes

2008 ◽  
Author(s):  
Shripad T. Revankar ◽  
Brian Wolf ◽  
Jovica R. Riznic
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Probabilistic Models ◽  
Crack Opening ◽  
Critical Flow ◽  
Probabilistic Assessment ◽  
Conditional Probabilities ◽  
Critical Flow Rate ◽  
Dose Limits ◽  
Steam Generator Tubes

The Canadian Nuclear Safety Commission (CNSC) recently developed the CANTIA methodology for probabilistic assessment of inspection strategies for steam generator tubes. Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits as an approach used to steam generator tube fitness-for-service assessment has been increasingly used in recent years throughout the nuclear power industry. The ANL/CANTIA code predictions were systematically studied to evaluate the code capability to predict the leak rates through the flawed steam generator tubes. In this evaluation the code models on the crack opening area, the probabilistic models and the critical flow rate models were studied and their applicability to available experimental data base was examined.

Numerical Study of the Deposition of Radioactive Nucleoids (Aerosol) in PWR Steam Generator

2021 ◽  
Author(s):  
Muhammad Aadil ◽  
Rab Nawaz ◽  
Ajmal Shah ◽  
Kamran Rasheed Qureshi
Keyword(s):  
Steam Generator ◽  
Numerical Study ◽  
Deposition Efficiency ◽  
Decontamination Factor ◽  
Radioactive Nuclide ◽  
Discrete Phase Model ◽  
Velocity Magnitude ◽  
Discrete Phase ◽  
Steam Generator Tubes ◽  
Secondary Side

Abstract This research presents numerical study of deposition efficiency and decontamination factor of radioactive nuclide in steam generator tubes of a typical 325 MWe PWR. To find out the deposition of aerosol, the discrete phase model (DPM) has been used. The flow has been characterized as compressible, adiabatic, turbulent and wall bounded. When steam generator tube gets ruptured, the radioactive nuclides can escape from primary side and create a radioactive field in the secondary side. This can be harmful for the personnel working at the plant. Therefore, in order to ensure the safety of the plant and personnel, it is important to study the particles deposition on the wall of steam generator tubes. In the present study, a CFD methodology has been first developed and validated with the published results. After methodology validation, it has been applied to the U-tube of a typical PWR steam generator. It has been observed that due to the action of centrifugal force near the bent, the velocity magnitude is high towards the inner wall and the flow separates at the bent entrance. Furthermore, the flow inside the tube is rotational with vortices throughout the domain due to the presence of the bent. Finally, the deposition efficiency and decontamination factor have been calculated and it has been observed that both increase with the increase in particle size due to inertial effects.

OXIPROBE: A Non Destructive Tool for Determining Steam Generator Oxide Characteristics

2009 ◽  
Author(s):  
John P. Krasznai
Keyword(s):  
Heat Transfer ◽  
Carbon Steel ◽  
Steam Generator ◽  
Transfer Efficiency ◽  
Steam Generators ◽  
Heat Transfer Efficiency ◽  
Tube Removal ◽  
Steam Generator Tubes ◽  
Non Destructive ◽  
Secondary Side

CANDU Stations are designed with significant amounts of carbon steel piping in the primary circuit. Although the primary coolant chemistry is such that carbon steel corrosion is minimized, nevertheless magnetite transport from the carbon steel surfaces to the steam generators is a significant issue leading to potential reduction in heat transfer efficiency in the steam generator. There are other contributors to the reduction of heat transfer efficiency such as divider plate leakage whereby some of the coolant short circuits the steam generator tubes and secondary side steam generator tube fouling. CANDU station operators have utilized a number of mitigating measures such as primary and secondary side mechanical and chemical tube cleaning, and divider plate refurbishment to counter these problems but these are all expensive and dose intensive, It is therefore very important to establish the relative contribution of each source to the overall heat transfer degradation problem so the most effective results are obtained. Tube removal and laboratory assessment of the oxide loading is possible and has been utilized but at best it provides an incomplete picture since typically only short lengths of tubes are removed — most often from the hot leg and the tube removal process adversely impacts the primary side oxide integrity. Kinectrics Inc. has developed, qualified and deployed Oxiprobe, a highly mobile non destructive technology able to remove and quantify the deposited oxide loading on the primary surfaces of steam generator tubes. The technology is deployed during shutdown and provides valuable, direct information on: • Primary oxide distribution within the steam generator; • Oxide loading (thickness of oxide) on the primary surfaces of steam generator tubes; • Oxide composition and radiochemical characterization. The End Effector probe can reach either side of the straight section of the steam generator U tube but as currently designed it is unable to be deployed in the U-tube region. The current technology is able to visit 4 tubes simultaneously. The technology is Code classified as a Class 6 fitting by the Canadian Nuclear Safety Commission and registered by the Ontario Technical Standards and Safety Authority as a pressure boundary retaining system. Although the application of the technology to date has been applied to steam generator tubes, in principle it can be applied to any heat exchanger tube, vertical or horizontal. This paper will describe the system, the qualification program for its deployment as well as some actual field results. The applicability of the technology for PWR steam generators is also addressed.

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.

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