Current Forgings and their Properties for Steam Generator of Nuclear Plant

2009 ◽  
pp. 56-56-9 ◽  
Author(s):  
H Tsukada ◽  
K Suzuki ◽  
M Kusuhashi ◽  
I Sato
Keyword(s):  
Steam Generator ◽  
Nuclear Plant

Influence of Steam Generator Tube Bundle Vibrations on the Operating Diagram of a Nuclear Plant During Stretch Out

2006 ◽  
Author(s):  
Andre´ Adobes ◽  
Joe¨l Pillet ◽  
Franck David ◽  
Michae¨l Gaudin
Keyword(s):  
Steam Generator ◽  
Cooling System ◽  
Tube Bundle ◽  
Thermal Power ◽  
Nuclear Plant ◽  
Coolant Temperature ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Normal Cycle ◽  
System Pressure

During the normal cycle of a pressurized water reactor, boron concentration is reduced in the core until fuel burns up. A stretch out of the normal cycle is however possible afterwards, provided primary coolant temperature is reduced. In those stretch out periods, nuclear operators want to keep constant thermal power exchanged in the steam generator, in order to preserve its performances. Under that constraint, the required reduction in primary coolant temperature involves both a decrease of secondary cooling system pressure and an increase of tube bundle vibrations. Since neither pressure nor vibrations should exceed some given thresholds in order to preserve component integrity, the reduction of primary coolant temperature has to be limited. Nuclear plant operators thereafter need an operating diagram, i.e. a diagram that provides minimum allowed primary coolant temperature versus power rate. In that context, we propose a method to derive such a diagram, by combining, on the one hand a code for simulating primary and secondary fluid flows in steam generators and, on the other hand, a software that allows one to predict fluid elastic tube bundle instabilities. That method allows one to take into account both tube fouling and plugging. It is now used by French utility “Electricite´ De France”, in order to check or supplement the analysis that are provided by steam generator manufacturers.

Flaw Growth Prediction and Fitness-for-Service Assessment of a PWR Nuclear Plant Steam Generator U-Tube

2002 ◽  
Author(s):  
Robert S. Vecchio ◽  
Sri K. Sam Sinha
Keyword(s):  
Finite Element ◽  
Fracture Mechanics ◽  
Crack Growth ◽  
Steam Generator ◽  
Degradation Mechanism ◽  
Element Model ◽  
Nuclear Regulatory Commission ◽  
Nuclear Plant ◽  
Pressurized Water ◽  
Nondestructive Examination

Primary water stress corrosion cracking (PWSCC) continues to be a dominant degradation mechanism affecting the service life of steam generators in several operating pressurized water reactor (PWR) nuclear plants. Recently, in one operating nuclear plant, a steam generator U-tube ruptured catastrophically while the unit was on-line. Although the plant operators were able to shutdown the reactor without significant release of radiation, the Nuclear Regulatory Commission (NRC) and the owner utility launched a full-scale investigation of the incident. The owner utility requested that a crack growth analysis and engineering evaluation of the tube rupture be performed, as well as assess the fitness-for-service of the generator for continued operation. This paper presents a summary of elastic-plastic finite element and fracture mechanics analyses performed for a steam generator U-tube, subjected to crack initiation at the inner diameter of the tube in the apex region. Residual stresses were computed from a finite element model of the tube simulating the mechanical bending process with the use of an anvil. Fracture mechanics and crack growth evaluations were performed to predict the time required for a pre-existing flaw at the inside diameter of the tube to propagate through-wall. Additionally, a fitness-for service assessment was performed in order to permit a degraded tube to remain in service, given an initial flaw size as determined by nondestructive examination.

Mathematical model of steam generator feed system at power unit of nuclear plant

2011 ◽  
Vol 72 (5) ◽  
pp. 1118-1126 ◽  
Author(s):  
E. M. Raskin ◽  
L. A. Denisova ◽  
V. P. Sinitsyn ◽  
Yu. V. Nesterov
Keyword(s):  
Mathematical Model ◽  
Power Unit ◽  
Steam Generator ◽  
Nuclear Plant ◽  
Feed System

NUMERICAL SIMULATION OF DRYOUT CHARACTERISTICS OF THE HELICAL ONCE-THROUGH STEAM GENERATOR

2018 ◽  
Author(s):  
Xiang Yu ◽  
Baozhi Sun ◽  
Jianxin Shi ◽  
Wanze Wu ◽  
Zhirui Zhao
Keyword(s):  
Numerical Simulation ◽  
Steam Generator
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