• Nuclear Engineering and Technology
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• 제36권1호
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• pp.53-63
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• 2004

The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same($0.008\%$), which showed that the CCMS could monitor core behaviors well.

• Nuclear Engineering and Technology
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• 제31권4호
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• pp.391-400
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• 1999

A conceptual core design of the 1,300MWe KNGR (Korean Next Generation Reactor) without using soluble boron for reactivity control was developed to determine whether it is technically feasible to implement SBF (Soluble Boron Free) operation. Based on the borated KNGR core design, the fuel assembly and control rod configuration were modified for extensive use of burnable poison rods and control rods. A new fuel rod, in which Pu-238 had been substituted for a small amount of U-238 in fuel composition, was introduced to assist the reactivity control by burnable poison rods. Since Pu-238 has a considerably large thermal neutron capture cross section, the new fuel assembly showed good reactivity suppression capability throughout the entire cycle turnup, especially at BOC (Beginning of Cycle). Moreover, relatively uniform control of power distribution was possible since the new fuel assemblies were loaded throughout the core. In this study, core excess reactivity was limited to 2.0 %$\delta$$\rho$ for the minimal use of control rods. The analysis results of the SBF KNGR core showed that axial power distribution control can be achieved by using the simplest zoning scheme of the fuel assembly Furthermore, the sufficient shutdown margin and the stability against axial xenon oscillations were secured in this SBF core. It is, therefore, concluded that a SBF operation is technically feasible for a large sized LWR (Light Water Reactor).

• Nuclear Engineering and Technology
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• 제14권2호
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• pp.51-63
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• 1982

한국에너지연구소에서 개발한 1차원적 제논과도현상해석 코드 DD1D를 사용하여 가압경수로심의 축방향 제논진동에 대한 안정성을 조사하였다. 노심의 출력준위, 감속재온도계수, 노심 입구온도, 도플러출력 계수 그리고 연소도의 변화가 노심의 축방향 안정성에 미치는 효과를 조사하기 위하여 고리1호기의 설계 및 운전자료를 이용하였으며 본 민감도 분석을 통하여 고리 1호기의 노심은 주기 초에는 축방향 제논진동에 대하여 안정하나 연소도가 증가함에 따라 안정도가 차츰 감소하여 주기 말에는 불안정해진다는 것을 알았다. 이같이 연소도가 증가함에 따라 노심의 안정도가 감소하는 이유는 연소도 변화에 따라 축방향의 출력분포, 감속재온도 계수 및 도플러출력계수가 변하기 때문이다. 본 연구를 통하여 출력밀도가 높은 대형 가압 경수로의 경우 전 주기동안 축방향제논진동에 대하여 안정된 노심을 설계하기 힘들다는 결론에 도달하였다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2002년도 하계학술대회 논문집 D
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• pp.2543-2545
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• 2002

There are many applicable services such as remote metering, load control, distribution line automation, Pole transformer Monitoring having their own networks in the electric power company. The application of the optical network technology as the back-born network to the Automation Systems in KEPCO is potentially beneficial in reliability, speed, and expandability. The 1-core and 2-core optical modems were developed and used by the Distribution Automation System. But, They had some disadvantages and advantages. So, We designed the new optical modems applied each advantages. This paper presents some of design efforts and test results for the multi-channel optical modem at KRPRI.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 춘계공동학술발표회요약집
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• pp.50-50
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• 2002

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3085-3094
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• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2321-2328
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• 2022

The passive safety system is adopted in an AP1000 nuclear power unit to improve the operation safety of the whole unit. However, due to boron diffusion and periodic sampling, boron dilution occurs in the core makeup tank. The boron replenishment process in the core makeup tank is essential and becomes particularly important. Based on the validated models, this article numerically investigates the influence of the replenishment flow rate and the position on the boron distribution in the core makeup tank. The thermal fatigue phenomenon of the "T" connection caused by replenishment is analyzed. Finally, the replenishment strategy is proposed to benefit both boron mixing in the core makeup tank and eliminating the thermal fatigue of the "T" connection.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.708-720
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• 2020

Core-wide temperature distribution in sodium-cooled fast reactor plays a key role in its decay heat removal process, however the prediction for temperature distribution is quite complex due to the conjugate heat transfer between the assembly flow and the inter-wrapper flow. Hybrid medium model has been proposed for conjugate heat transfer modeling in the core. The core is modeled with a Realistic modeled inter-wrapper flow and hybrid medium modeled assembly flow. To validate present model, simulations for a three-assembly model were performed with Realistic modeling, traditional porous medium model and hybrid medium model, respectively. The influences of Uniform/Non-Uniform power distribution among assemblies and the Peclet number within the assembly flow have been considered. Compared to traditional porous medium model, present model shows a better agreement with in Realistic modeling prediction of the temperature distribution and the radial heat transfer between the inter-wrapper flow and the assembly flow.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.97-106
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• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.