• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.204-204
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• 2002

• Journal of Magnetics
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• v.20 no.4
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• pp.439-443
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• 2015

A control rod drive mechanism (CRDM) is an electromechanical equipment that provides linear movement for the control rods to control the nuclear reactivity in a nuclear reactor. A rod position indication system (RPIS) detects the control rod's position. To enhance the measurement accuracy of the system, a magnetostrictive type sensor with capability of generating operation limiting signals would be adapted instead of a conventional RPIS for a CRDM. An RPIS was modelled for a numerical analysis with the permanent magnets at the stationary limit positions and magnetic shielding plates with a moving permanent magnet. The performance analysis of the RPIS were conducted, and the results were discussed here.

• Proceedings of the KIEE Conference
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• 2002.07d
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• pp.2295-2297
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• 2002

In this paper, we describe a set of test facilities for the development of CRCS (Control Rod Control System). The test facilities consist of a code simulator CRDM (Control Rod Drive Mechanism) mockup, an input/output simulator for validation work and R-L load to simulate CRDM mockup.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.258-263
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• 1998

원자력 발전소의 제어봉 낙하 시간 측정 시험은 원자로 비상 정지 명령에 따른 제어봉의 낙하 속응성이 지침서의 제한치 이내에 있는지를 주기적으로 확인함으로써 제어봉의 원자로 안전 운전 및 정지 기능을 확인하는 중요한 시험이다. 현재 고리2호기를 비롯한 WH형 발전소들은 제어봉이 디지틀 제어봉 위치 지시(DRPI) 코일속을 낙하할 때 유기되는 전압을 검출하여 제어봉 낙하 시간을 측정하는 방법을 사용하고 있다. 이 방법은 제어봉을 뱅크별로 인출한 후 하나씩 개별 낙하시키기 때문에, 제어봉 낙하 시간 측정 시험에 많은 시간과 다수의 인력이 소요되며, 측정시 격납용기내에 작업자가 장시간 체류해야 하는 단점이 있다. 이러한 DRPI 신호 이용 제어봉 낙하 시간 측정 방법의 단점을 보완하여 제어봉 구동 장치(CRDM)내 이동권선 신호를 이용하는 다중 낙하 방식의 새로운 측정 방법을 제안하고, 측정 장치를 개발하였다. 개발된 제어봉 낙하 시간 측정 장치를 이용하여, 고리2호기 제어봉 낙하 시간 측정 시험시 기존의 방법과 병행 측정 시험을 수행하고 그 결과를 비교 검토하여 충분히 정확하게 제어봉 낙하 시간을 측정할 수 있음을 보였다.

• Proceedings of the KIEE Conference
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• 2007.10a
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• pp.291-292
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• 2007

This paper deals with the design of control cabinet based on safety PLC for Control Rod Control System(CRCS). The CRCS controls the operation of the CRDMs(Control Rod Drive Mechanisms). The CRDM moves the control rods which regulate the reactor power. vertically in the reactor core. The Control Cabinet in CRCS makes and conveys control signals to the power cabinet which provides power to the CRDM. We designed the Control Cabinet, based on POSAFE-Q, safety PLC. The application programs working in PLC can be programmed by pSET(POSAFE-Q Software Engineering Tool), Identified Development Environment.

• Proceedings of the KIEE Conference
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• 2003.11c
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• pp.575-579
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• 2003

In this paper, we describe a Duplex Power Controller for Control Rod Control System(CRCS). A Duplex Power Controller has the various functions for the reliable operations of Control Rod Drive Mechanism(CRDM). Also we have implemented the diverse functions by utilizing the developed Duplex Power Controller. Due to the developed Duplex Power Controller, we are assured that the commercial operation by this system be made before long.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.345.1-345
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• 1999

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.989-990
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.967-968
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• 2005