Design and development of robotic automation and process controls for the water jet lancing system of nuclear power plant steam generators

2006 ◽  
Author(s):  
J Aparna ◽  
B.S.V.G Sharma
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water Jet ◽  
Steam Generators ◽  
Design And Development ◽  
Robotic Automation ◽  
Process Controls

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.

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.

Operating Condition of Steam Injector as a Passive Cooling System at Severe Accident of Nuclear Power Plant

2012 ◽  
Author(s):  
Yutaka Abe ◽  
Shunsuke Shibayama ◽  
Akiko Kaneko ◽  
Chikako Iwaki ◽  
Tadashi Narabayashi ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Structure ◽  
Test Section ◽  
Nuclear Power ◽  
Water Jet ◽  
Back Pressure ◽  
Severe Accident ◽  
Opening Ratio ◽  
Discharge Pressure

Steam injector (SI) is a passive jet pump which is driven by high-performance steam condensation onto water jet and it is expected to be active at severe accident of nuclear power plant with no electricity. SI is mainly consists of convergent-divergent nozzle. Supersonic steam flow condenses onto water jet in the mixing nozzle and mass, momentum, and energy of steam is transferred to water in the mixing nozzle. Condensed water jet is accelerated at the throat and kinetic energy is converted into pressure in the diffuser, which produces higher pressure than inlet steam pressure. It is easy to apply the SI to nuclear power plant since SI has quite simple and compact structures. The objectives of the present study are to clarify the mechanism of heat and momentum transfer in the mixing nozzle and to determine operating range of SI for practical use. A transparent test section is adopted to conduct visualization of the flow structure with a high-speed video camera as well as measurement of pressure distribution in mixing nozzle, throat, and diffuser with changing back pressure. Fundamental parameters change between operative and inoperative state of the injector were evaluated by measuring pressure and temperature distribution along axial direction of the test section. Discharge pressure as one of operating characteristics of the injector was also measured in changing back pressure by decreasing the opening ratio of the back pressure valve attached downstream of the test section. It was confirmed that discharge pressure increased and the injector became inoperative unsteadily with decreasing opening ratio of the back pressure valve just after it produced the maximum discharge pressure. In the present investigation, this maximum discharge pressure is evaluated as the operation limit of the injector. Furthermore, discharge pressure from diffuser, which is one of the indicators of operating performance as well as operating limit is predicted from inlet condition adopting one-dimensional analysis model proposed previously. By comparing analytical result with experimental data, as well as visualization of flow structure in throat and diffuser, physics model including two-phase flow structure with shock wave which was observed at throat and diffuser are discussed in order to predict injector’s operation with high accuracy.

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