• Journal of Energy Engineering
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• v.24 no.2
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• pp.91-102
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• 2015

A thermal-hydraulic design and performance analysis computer code for a once-through steam generator using straight tubes is developed. To benchmark the developed physical models and computer code, an once-through steam generator developed by other designer is simulated and the calculated results are compared with the design data. Also, the same steam generator is analyzed with the best-estimate thermal-hydraulic system code, MARS, for the code-to-code validation. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by the developed code show general agreements with the published design data as well as the analysis results of MARS. It is demonstrated that the developed code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.125-134
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• 2009

This paper presents details on a tube inspection robotic system and a positioning method of the robot for a steam generator (SG) in nuclear power plants (NPPs). The robotic system is separated into three parts for easy handling, which reduces the radiation exposure during installation. The system has a supporting leg to increase the rigidity of the robot base. Since there are several thousands of tubes to be inspected inside a SG, it is very important to position the tool of the robot at the right tubes even if the robot base is positioned inaccurately during the installation. In order to obtain absolute accuracy of a position, the robot kinematics was mathematically modeled with the modified DH(Denavit-Hartenberg) model and calibrated on site using tube holes as calibration points. To tune the PID gains of a commercial motor driver systematically, the time delay control (TDC) based gain tuning method was adopted. To verify the performance of the robotic system, experiments on a Framatomes 51B Model type SG mockup were undertaken.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.859-866
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• 2001

During the Pre-service Inspection for Wolsong Unit 3&4 in 1997/1998 respectively, 17 Distorted Roll Transition indications(over expanded beyond tubesheet secondary face) were identified at the Unit 4 (S/G B, D). Six(6) tubes out of these tubes were plugged in 1998. However the first Periodic Inspection identified additional 110 indications in 1999 and 2000. The additionally identified 110 indication call, not reported at the Pre-service Inspection, are; 2 Not-Finally-Expanded-Tubes and 108 Distorted Roll Transition tubes. Design limit of each Steam Generator tube Plugging is 6.4%. Plugging was performed by the Steam Generator manufacturer under the warranty. When Distorted Roll Transition indications were first identified on the Unit 4 in 1998 the degree of Over-expansion was measured using an inner dial-gage to make the disposition of Nonconformance report. 2 Not-Finally-Expanded-Tubes were plugged and 10 tubes out of 108 Distorted Roll Transition Tubes were also plugged as a preventive measure.

• Journal of the Korean Institute of Intelligent Systems
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• v.16 no.2
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• pp.191-197
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• 2006

This paper deals with a fault detection system design for nuclear steam generator water level control system. We expressed the nonlinear properties of the steam generator level system as a T-S fuzzy system with time varying uncertain parameters. We design a residual generator using a left coprime factorization of the T-S fuzzy model and a fault detection filter in order to improve the fault detection performance. We demonstrate the efficiency of the suggested design method via many computer simulations.

• Journal of Electrical Engineering and information Science
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• v.2 no.3
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• pp.66-70
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• 1997

The steam generator level is susceptible to the nonminimum phase in dynamics due to the thermal reverse effects known as "shrink and swell" in a pressurized water reactor. A state feedback assisted control concept is presented for the change of dynamic performance to the minimum phase the concept incorporates a nonlinear digital observer as a part of the control system. The observer is deviced to estimate the state variables that provide the true indication of water inventory by compensating for shrink and swell effects. The concept is validated with implementation into the steam generator simulation model.

• Proceedings of the KIEE Conference
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• 1993.07a
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• pp.265-267
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• 1993

A control for Steam Generator (S/C) level is very difficult by automatic control mode but also manual control node during plant start up and/or low power level operation with the analog control system because of a non-nominal process responce. The goal of this study is to improve and computerize and applicate for KO-RI #1 Steam Generator level control system.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.501-509
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• 2001

For the enhancement of ECT reliability on the primary water stress corrosion cracks of nuclear steam generator tubes, of which the occurrence is on the increase, it is important to comprehend the signal characteristics on crack morphology and to select an appropriate probe type. In this paper, the sizing accuracy and the detectability for the inner wall axial cracks of tubes were quantitatively evaluated using the following specimens: the electric discharge machined notches and the corrosion cracks which were developed on the operating steam generator tubes. The difference of eddy current signal characteristics between pancake and axial coil were also Investigated. The results obtained from this study provide a useful information for more precise evaluation on the inner wall axial tracks oi stram generator tubes.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.6
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• pp.519-523
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• 2013

A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.833-844
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• 1995

In self-tuning predictive control algorithm for steam generator is presented. The control algorithm is derived by suitably modifying the generalized predictive control algorithm. The main feature of the unposed method relies on considering the measurable disturbance and a simple adaptive scheme for obtaining the controller gain when the parameters of the plant are unknown. This feature makes the proposed approach particularly appealing for water level control of steam generator when measurable disturbance is used. In order to evaluate the performance of the proposed algorithm, computer simulations are done for an PWR steam generator model. Simulation result show satisfactory performances against load variations and steam flow rate estimation errors. It can be also observed that the proposed algorithm exhibit better responses than a conventional PI controller.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.23 no.7
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• pp.67-74
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• 2009

This paper describes the speed control functions of the typical steam turbine speed controllers and the test results of generator load rejection simulations. The goal of the test is to verify the speed controller's ability to limit the steam turbine's peak speed within a predetermined level in the event of generator load loss. During normal operations, the balance between the driving force of the steam turbine and the braking force of the generator load is maintained and the speed of the turbine-generator is constant. Upon the generator's load loss, in other word, the load rejection, the turbine speed would rapidly increase up to the peak speed at a fast acceleration rate. It is required that the speed controller has the ability to limit the peak speed below the overspeed trip point, which is typically 110[%] of rated speed. If an actual load rejection occurs, a substantial amount of stresses will be applied to the turbine as well as other equipments, In order to avoid this unwanted situation, not an actual test but the other method is necessary. We are currently developing the turbine control system for another nuclear power plant and have plan to do the simulation suggested in this paper.