• 전산구조공학
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• 제3권3호
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• pp.141-146
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• 1990

본 연구에서는 다목적연구로 반응도제어장치의 제어봉에 대한 내진해석을 수행하였다. 해석모델은 물속에 잠겨있는 두개의 관(중성자흡수봉 및 유동관)이 동심축상에 있으며 부분적으로 중첩된 두 관끝단의 유체틈에 의해 서로 동적으로 연결되어 있다. 유체에 의한 동적질량을 고려한 고유진동수를 구하고 안전정지지진에 의해 발생하는 최대응력과 최대변위를 동위상 및 역위상 거동별로 각각 구하였다. 해석결과 최대응력은 허용치보다 작게 나타나 구조적인 건전성은 입증되었으나, 최대변위는 두 개의 관이 서로 부딪히고 중성자흡수봉은 주위의 다른 벽에도 부딪히는 현상을 수발함을 알 수 있었다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2003년도 하계학술대회 논문집 D
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• pp.2271-2273
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• 2003

Control Rod Control System(CRCS) is to control nuclear reaction of reactor by moving Control Rod Drive Mechanism(CRDM) with speed and direction signal from Reactor Regulating System(RRS). CRCS is made up of two parts : control cabinet and power cabinet. And this paper presents mainly power cabinet design for system reliability. To increase reliability of power cabinet, controller, power supply and communication line arc doubly designed and supervision and diagnosis function are applied.

• 한국복합재료학회:학술대회논문집
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• 한국복합재료학회 2001년도 춘계학술발표대회 논문집
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• pp.159-162
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• 2001

A combined finite element and experimental study based on the characteristic length method is performed to investigate the strength and behavior of the pin joint in composite control rod. The failure is estimated by the Yamada-Sun and Tsai-Wu criteria on the characteristic curve. The gap elements are used to simulate the contact between the pin and the composite fitting with hole. The accuracy and applicability of the method are validated by the joint tests. All the specimens were failed in the bearing mode in the test and finite element analysis, and good agreement was found between the predicted and test results on the joint strength of composite control rod.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2006년 학술대회 논문집 정보 및 제어부문
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• pp.269-272
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• 2006

In this paper we describe the design of a Dual Power Controller(DPC) for Power Control System(PCS) in Nuclear Power Plant. The PCS also provides information regarding rod motion, rod position, and status of the Rod Control System. It has Hot/Stand-by type, and also has the function of fault detection for controller itself and power modules. We have implemented the various functions with the dual Power Controller. Due to the developed DPC, we are assured that the commmecial use by this controller be made before long.

• Nuclear Engineering and Technology
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• 제31권6호
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• pp.1-7
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• 1999

An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system(SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet(CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid MEtal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1834-1846
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• 2020

The SMART (System-integrated Modular Advanced ReacTor) is an integral-type small modular reactor developed by KAERI (Korea Atomic Energy Research Institute). This paper discusses the development of a dynamic equivalent model of the SMART control rod drive mechanism that can be efficiently utilized for complicated analysis during the design of the SMART. A semi-empirical approach is used to develop the equivalent model; that is, the equivalent model is defined analytically and verified empirically. Two types of tests, dynamic characteristics tests and seismic loading tests, are conducted for the development and verification of the dynamic equivalent model, respectively. Acceleration response spectra from the seismic analysis based on the developed equivalent model show good agreement with those from the seismic loading tests.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.80-87
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• 1999

An analytic axial xenon oscillation model was developed for pressurized water reactor analysis. The model employs an equation system for axial difference parameters that was derived from the two-group one-dimensional diffusion equation with control rod modeling and coupled with xenon and iodine balance equations. The spatial distributions of nu, xenon, and iodine were expanded by the Fourier sine series, resulting in cancellation of the flux-xenon coupled non-linearity. An inhomogeneous differential equation system for the axial difference parameters, which gives the relationship between power, iodine and xenon axial differences in the case of control rod movement, was derived and solved analytically. The analytic solution of the axial difference parameters can directly provide with the variation of axial power difference during xenon oscillation. The accuracy of the model is verified by benchmark calculations with one-dimensional reference core calculations.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 1990년도 봄 학술발표회 논문집
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• pp.44-49
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• 1990

This study is a seismic analysia of absorber rod in KMRR Reactivity Control Mechanism. The model being studied i8 two coaxial tubes (control absorber rod and flow tube) immersed in the water and partially coupled (overlap) by water Hap. The hydrodynamic mass effects by the water in each surrounding conditions are considered in the model. The natural frequencies, stresses and displacements of the system due to Safe Shutdown Earthquake are computed in the cases of in-phase modes and out-of-phase modes of two coaxial tubes. The results show that maximum stresses are well below the allowable limit and maximum displacements at the ends of both tubes in out-of-phase modes are so huck that the tubes contact each other in the overlap zone.

• Nuclear Engineering and Technology
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• 제30권4호
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• pp.342-352
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• 1998

A nuclear design feasibility of soluble boron free(SBF core for the medium-sized(600MWe) PWR was investigated. The result conformed that soluble boron free operation could be performed by using current PWR proven technologies. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with burnable poison and control rod absorber material. In order to control excess reactivity, large amount of gadolinia integral burnable poison rods were used and B4C was used as a control rod absorber material. For control of bottom shift axial power shape due to high temperature feedback in SBF core, axial zoning of burnable poison was applied to the fuel assemblies design. The combination of enrichment and rod number zoning for burnable poison could make an excess reactivity swing flat within around 1% and these also led effective control on axial power offset and peak pin power, The safety assessment of the designed core was peformed by the calculation of MTC, FTC and shutdown margin. MTC in designed SBF core was greater around 6 times than one of Ulchin unit 3&4. Utilization of enriched BIO(up to 50w1o) in B4C shutdown control rods provided enough shutdown margin as well as subcriticality at cold refueling condition.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2004년도 학술대회 논문집 정보 및 제어부문
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• pp.559-561
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• 2004

In this paper, we design two types of Main Control Unit for Control Rod Control System and Control Element Drive Mechanism Control System, respectively, using a domestic Distributed Control System(DCS) developed to localize the instrumentation and control(I&C) system for nuclear power plant(NPP). There are many parts developed by domestic skills and being operated successfully in NPP, but the development of I&C system as an essential part has been slow in progress. We will show the great possibility of developing peculiar Korean I&C system by applying this domestic DCS to nuclear I&C system and confirming its successful operation.