scholarly journals Analysis of Pipe Wall-thinning Caused by Water Chemistry Change in Secondary System of Nuclear Power Plant

2015 ◽  
Vol 14 (6) ◽  
pp. 325-330 ◽  
Author(s):  
Hun Yun ◽  
Kyeongmo Hwang ◽  
Seung-Jae Moon
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Chemistry ◽  
Nuclear Power ◽  
Pipe Wall ◽  
Secondary System ◽  
Wall Thinning

A Feasibility Study of Noncontact Ultrasonic Sensor for Nuclear Power Plant Inspection

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Akinori Tamura ◽  
Chenghuan Zhong ◽  
Anthony J. Croxford ◽  
Paul D. Wilcox
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Thickness Measurement ◽  
Piping System ◽  
Wall Thinning ◽  
Plant Inspection ◽  
Ultrasonic Thickness

A pipe-wall thinning measurement is a key inspection to ensure the integrity of the piping system in nuclear power plants. To monitor the integrity of the piping system, a number of ultrasonic thickness measurements are manually performed during the outage of the nuclear power plant. Since most of the pipes are covered with an insulator, removing the insulator is necessary for the ultrasonic thickness measurement. Noncontact ultrasonic sensors enable ultrasonic thickness inspection without removing the insulator. This leads to reduction of the inspection time and reduced radiation exposure of the inspector. The inductively-coupled transducer system (ICTS) is a noncontact ultrasonic sensor system which uses electromagnetic induction between coils to drive an installed transducer. In this study, we investigated the applicability of an innovative ICTS developed at the University of Bristol to nuclear power plant inspection, particularly pipe-wall thinning inspection. The following experiments were performed using ICTS: thickness measurement performance, the effect of the coil separation, the effect of the insulator, the effect of different inspection materials, the radiation tolerance, and the measurement accuracy of wastage defects. These initial experimental results showed that the ICTS has the possibility to enable wall-thinning inspection in nuclear power plants without removing the insulator. Future work will address the issue of measuring wall-thinning in more complex pipework geometries and at elevated temperatures.

scholarly journals Analysis of Wall-Thinning Effects Caused by Power Uprates in the Secondary System of a Nuclear Power Plant

2016 ◽  
Vol 15 (3) ◽  
pp. 135-140
Author(s):  
Hun Yun ◽  
Kyeongmo Hwang ◽  
Hyoseoung Lee ◽  
Seung-Jae Moon
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Secondary System ◽  
Wall Thinning

Water chemistry effect on the wear of stainless steel in nuclear power plant

2006 ◽  
Vol 39 (12) ◽  
pp. 1503-1508 ◽  
Author(s):  
D. Kaczorowski ◽  
P. Combrade ◽  
J.Ph. Vernot ◽  
A. Beaudouin ◽  
C. Crenn
Keyword(s):  
Stainless Steel ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Water Chemistry ◽  
Nuclear Power

Mechanical and Hydraulic Loads on the Supports of the Rotary Machine Due to Shaft Seizure in a Nuclear Power Plant

2004 ◽  
Author(s):  
S. K. Gupta ◽  
B. Chatterjee ◽  
Rajesh Kumar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Transport System ◽  
Nuclear Power ◽  
Inertial Load ◽  
Secondary System ◽  
Rotating Shaft ◽  
Rotary Machine ◽  
Major Equipment

In a nuclear power plant there are two major equipment with high mechanical inertia, which have a rotating shaft. These are Pumps in the Primary Heat Transport System and Turbine in the secondary system. In both cases, the shaft seizure leads to transfer of very large loads to the supports. These supports, if not designed for seizure loads may fail. If the supports fail, there is a good possibility of a missile generated and hit the safety equipment. Seizure loads in these machines have three components namely mechanical inertial load, electrical load and hydraulic load. While the electrical and hydraulic loads have a limited peak value, the inertial load depends on the seizure time. For the normal observed seizures the three have a similar order of magnitude during seizure. As the casing is overdesigned the combined load is experienced by the supports. The pump of the Primary Heat Transport System (PHTS) of a nuclear power plant is centrifugal type run by an induction motor. If the pump shaft seizes, the seizure load will be experienced by the support structure. Due to the presence of the flywheel, the total moment of inertia of the pump motor assembly is quite high. Hence the resisting torque may be higher than the support’s design torque. Besides, the electric torque will continue to be applied as the motor trip on the overload current is delayed by several seconds as the corresponding relay is a thermal relay. Seizure torque will depend on pump seizure time. Lesser the seizure time, higher would be the load on the pump supports. The turbine in the secondary system has a large inertia due to blades. In case of a seizure the generator is tripped in hundreds of milliseconds. The load experienced by turbine supports due to seizure is significantly enhanced in the first few seconds due to sustained steam supply before it is cut off. This paper discusses the estimation of the three types of loads during seizure of the shafts in the pumps and turbine. It also discusses the possible safety consequences of these loads.

Coupling of RELAP and LabVIEW to Model a Nuclear Power Plant System Realistically

2014 ◽  
Author(s):  
J. C. Pack ◽  
Z. Fu ◽  
F. Aydogan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Virtual Instrument ◽  
Nuclear Power ◽  
System Modeling ◽  
Loop Model ◽  
Secondary System ◽  
Loss Of Coolant Accident ◽  
Accident Scenarios ◽  
Computer Simulators

Within the study and design of a nuclear power plant extensive system modeling is necessary to determine how the reactor is going to perform in any given situation, not only in the normal performance of the reactor but also transients including unanticipated transients without scram and hypothetical accidents. One of the difficulties in the performance of this modeling is that there are often separate programs used to model the primary and other loops in multiple loop systems. When the modeling requires no interaction between the loops, this method is adequate but in many of these scenarios an understanding of the interaction loops is crucial especially in the case of transients including accident scenarios. However, each loop is generally modeled individually and there is no feedback effect between loops. The purpose of this article is to demonstrate how this coupling between the primary and secondary system of a typical PWR can be performed. The primary and secondary sides of the PWR are modeled with Reactor Excursion and Leak Analysis Program (RELAP5) and Laboratory Virtual Instrument Engineering Workbench (LabVIEW) computer simulators respectively. Primary loop model includes a four loops PWR. The coupling between RELAP5 and LabVIEW has been executed with steady state and transients, in this case a loss of coolant accident (LOCA). The results of the coupling have been compared with the typical RELAP5 results without coupling.

Dynamic Simulation on Secondary System of Nuclear Power Plant

2012 ◽  
Vol 591-593 ◽  
pp. 620-625 ◽  
Author(s):  
Rui Yu ◽  
Shi Wei Yao ◽  
Chun Guo Wang
Keyword(s):  
Steady State ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Dynamic Simulation ◽  
Nuclear Power ◽  
Large Scale ◽  
Simulation Models ◽  
Secondary System ◽  
Actual System ◽  
Moisture Separator

The secondary system of Qinshan phase I nuclear power plant is simulated in this study. According to the characteristics of the JTopmeret model, the system is divided into six parts for modeling, which are the deaerator, the high pressure (HP) turbine, the low pressure (LP) turbine, the moisture separator reheater (MSR), the condensate system, and the feedwater system. All parts are built as the graphic automatic models in JTopmeret and debugged on the large-scale simulation platform GSE to complete the steady-state and dynamic simulation of the models. The results show that the steady-state and dynamic processes of the models are consistent with the characteristics of the actual system. It verifies the correctness of the simulation models. Thus, this research is able to provide guidance for the operation analysis and the equipment debugging of the secondary system of the nuclear power plant.

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