• Transactions of the KSME C: Technology and Education
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• v.3 no.3
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• pp.225-232
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• 2015

The control element drive mechanism (CEDM) is an electro-mechanical device to control reactivity of the nuclear reactor. The conventional CEDM was installed on a nozzle of the reactor vessel closure head as an ex-vessel type. However, there have been demands for an in-vessel CEDM to fundamentally eliminate the rod ejection accident. Conceptual design of the in-vessel CEDM, which was developed based on the existing technology of the ex-vessel CEDM, is introduced in this paper.

• Proceedings of the KIEE Conference
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• 1995.07b
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• pp.761-765
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• 1995

The old Timing Controller for Control Element Drive Mechanism (CEDM) is designed as an open loop control system because it is difficult to mount sensors within the Control Element Drive Mechanism(CEDM) which is operating under the pressure boundary of the reactor vessel. In this work new method which can be used to detect the CEDM operational conditions without mounting sensors within the CEDM housing is developed in order to resolve problems of the old Timing Controller. By using the developed new method, the new Timing Controller for the CEDM is designed as a closed loop controller which has features of the control rod drop prevention, fine position control and the coil life time extension. The algorithm developed under closed loop control concept resolves most problems occurred in the old Timing Controller and improves the performance and reliability of the system. During designing and testing of the Timing Controller algorithm, the real time CEDM simulator developed here was used. And all functions of the developed algorithm were verified using CEDM simulator with the real data collected from the site. The results show that the Timing Controller performs its intended functions properly.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.637-647
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• 2014

In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1235-1239
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• 2012

Primary water stress corrosion cracking (PWSCC) has been observed around the weld region of control rod drive mechanism (CRDM) nozzles in nuclear power plants overseas. The weld has a J-shaped groove and it connects the CRDM nozzle with the reactor vessel upper head (RVUH). It is a dissimilar metal weld (DMW), because the CRDM is made of alloy 600 and the RVUH is made of carbon steel. In this study, finite element analysis (FEA) was performed to estimate the stress condition around the weld region. Generally, it is known that a high tensile region is more susceptible to PWSCC. FEA was performed as for the condition of welding, hydrostatic test and normal operation successively to observe how the residual stress changes due to plant condition. The FEA results show that a high tensile stress region is formed around the weld starting point on the inner surface and around the weld stop point on the outer surface.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1997.04a
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• pp.147-151
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• 1997

The dynamic characteristics of Control Element Drive Mechanism (CEDM) in Korea standard Nuclear Power Plant was reviewed as a secondary mass in a simplified two degree of freedom system, while the reactor vessel as a primary mass. The design improvement stratege to minimize each displacement amplitude of these primary and secondary masses was proposed. According to this stratege the designs of CEDM components, the shroud and the pressure housing, respectively, were changed using optimization technique.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.11
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• pp.1848-1853
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• 2016

Control Rod Position Indicator in nuclear reactor vessel has developed for small reactor in Korea. Because of severe environment in reactor vessel, target of this study is to develop the suitable position indicator. In this study, solenoid type position indicator made of Mineral Insulated Cable(MI Cable) was introduced to adapt in severe environment. And inductance of the solenoid was used to indicate the rod position for high precision. But problem of this concept is that a linear slope of inductance is changed by temperature effect. To resolve this problem, two sensing coils were introduced for temperature compensation. A role of the sensing coil is to make reference linear equation about certain temperature. To confirm this concept, also, inductance of solenoid and sensing coils were measured at room and high temperature (${\sim}300^{\circ}C$). The results of measurement show that the position error of sensing coil between room and high temperature was about 2%. But it was identified that this error was resulted from insufficient test environment (temperature error between solenoid and sensing coils was about 2% at high temperature condition). Therefore, solenoid type position indicator shows that it is very suitable in reactor vessel as a high precision rod position indicator.

• Computational Structural Engineering
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• v.10 no.3
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• pp.225-231
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• 1997

The dynamic characteristics of Control Element Drive Mechanism(CEDM) for Korea Standard Nuclear Power Plant are studied with the CEDM modeled as a secondary mass in a simplified two degree of freedom system, while the reactor vessel as a primary mass. The optimal .mu.-f curve is developed to reduce the response amplitudes of both primary and secondary masses. In order to improve a design it is proposed that the natural frequency ratio, f, should be converged to 0.93, the mass ratio, .mu., should not be reduced, and the result should be converged to the optimal .mu.-f curve. Optimal design for CEDM components has been carried out and the response amplitude ratios of reactor are reduced 10.5 - 19.7% while those of CEDM are reduced 6.3 - 3.4%.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.10
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• pp.1337-1345
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• 2011

In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor$r_o$/t, geometry of fillet, and adjacent nozzle.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1159-1168
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• 2013

In pressurized water nuclear reactors (PWRs), the reactor pressure vessel (RPV) upper head contains penetration nozzles that use a control rod drive mechanism (CRDM). The penetration nozzle uses J-groove weld geometry. Recently, the occurrence of cracking in alloy 600 CRDM penetration nozzle has increased. This is attributable to primary water stress corrosion cracking (PWSCC). PWSCC is known to be susceptible to the welding residual stress and operational stress. Generally, the tensile residual stress is the main factor contributing to crack growth. Therefore, this study investigates the effect on weld residual stress through different analysis methods for normal operating conditions using finite element analysis. In addition, this study also considers the effect of repeated normal operating condition cycles on the weld residual stress. Based on the analysis result, this paper presents a normal operating condition analysis method.