Using probabilistic methods for solving the problem of ensuring leak tightness of heat exchanger tubes of nuclear power plant steam generators

2012 ◽  
pp. 270-323
Author(s):  
Gennadij V. Arkadov ◽  
Alexander F. Getman ◽  
Andrei N. Rodionov
Keyword(s):  
Heat Exchanger ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Probabilistic Methods ◽  
Steam Generators ◽  
Leak Tightness

scholarly journals Heat transfer intensification in emergency cooling heat exchanger of nuclear power plant using air-water mist flow

2019 ◽  
Vol 2019 (3) ◽  
pp. 16-27
Author(s):  
Akram Hamzah Abed ◽  
Sergej Ewgen’evich Shcheklein ◽  
Valery Maksimovich Pakhaluev
Keyword(s):  
Heat Transfer ◽  
Heat Exchanger ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Mist ◽  
Heat Transfer Intensification ◽  
Mist Flow

Numerical model for estimation of corrosion location in nuclear power plant steam generators

2011 ◽  
Vol 241 (1) ◽  
pp. 95-100 ◽  
Author(s):  
S. Tashakor ◽  
G. Jahanfarnia ◽  
A. Kebriaee
Keyword(s):  
Power Plant ◽  
Numerical Model ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Steam Generators

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Research on the Inspection Strategy Change Based on Risk-Informed Method in Nuclear Power Plant

2017 ◽  
Author(s):  
Ye Shuixiang ◽  
Li Qiongzhe ◽  
Xi Haiying
Keyword(s):  
Heat Exchanger ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
System Component ◽  
Informed Decision Making ◽  
Inspection Strategy ◽  
Strategy Change ◽  
Generation Activity

In order to gain more operational benefits, many nuclear power plants will intend to optimize the generation activity to increase generating capability factor, such as, shortening the refueling schedule by optimize the inspection strategy of relevant to SSCs (structure, system, component) during reactor shutdown. Therefore, this paper will introduce the research on the inspection strategy change based on risk-informed method, and provide detailed discussion on change of the RHR heat exchanger inspection strategy by employed the above mentioned method in one nuclear power plant. The result shows that the optimization of RHR heat exchanger inspection strategy meets all the consideration of the principles in risk-informed decision-making required by RG 1.174 and the optimization proposal is acceptable.

Controlling and Supervising the Nuclear (Core) Power of a PWR in Relation to Nuclear Safety and Economics

2009 ◽  
Author(s):  
P. Wouters ◽  
W. Van Rompay ◽  
F. Bertels ◽  
W. Van Hove ◽  
E. Gorleer ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Total Error ◽  
Payback Period ◽  
Measuring System ◽  
Steam Generators ◽  
Nuclear Core ◽  
Ultrasonic Flow Meter

Knowing exactly the nuclear core power of a nuclear reactor is one of the most important parameters for the operator; it is vital for safety as well as for economical matters. The secondary calorimetric is the only one where one can pilot on; it is a combination of measured parameters, of which the feedwater (FW) flow towards the steam generators is the most significant one. This feedwater flow can be measured by means of an ultrasonic flow meter, “LEFM CheckPlus™ system” instead of the commonly used venturis or diaphragms. In the Belgian Nuclear Power Plant (NPP) Doel 4, a new ultrasonic “LEFM CheckPlus™” feedwater flow measuring system has been installed in April 2008. The paper describes the consequences of the installation, as the total error on the secondary calorimetric decreases from the previous 1,3% to the current 0,8% with a possibility of further reduction to 0,4%. Additionally, the economical effects of the installation are calculated for a 1000 MWe power plant with venturi meters undergoing fouling. For the NPP Doel 4 it was an economically interesting investment since the payback period was only 45 days. Finally, the possibility of consuming the margin on the secondary calorimetric for a mini-power uprate is inspected, technically and economically. It is concluded that such a mini-power uprate is an interesting option for the NPP owner.

scholarly journals Corrosion Monitoring of the Steam Generators of V-th and VI-th Energy Blocks of Nuclear Power Plant “Kozloduy”

10.5772/17186 ◽  
2011 ◽  
Author(s):  
Nikolai Boshkov ◽  
Georgi Raichevski ◽  
Katja Minkova ◽  
Penjo Penev
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Corrosion Monitoring ◽  
Steam Generators
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