• 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.

• 마이크로전자및패키징학회지
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• 제13권4호
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• pp.9-15
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• 2006

이 논문은 코어와 I/O 회로가 포함된 PEEC(Partial Equivalent Electrical Circuit) PDN(Power Distribution Networks)의 임피던스 변화에 따른 칩의 성능 분석을 나타내었다. I/O 전원에 연결된 코어 전원 잡음이 I/O 스위칭에 어떠한 영향이 미치는지 시뮬레이션 결과를 통하여 보였다. 또한 직접 설계한 $7{\times}5$인치 DLL(Delay Locked Loop)시험 보드를 사용하여 칩의 동작 지점에 따른 전원 잡음의 효과를 분석하였다. $50{\sim}400MHz$에 주파수 대역에 따른 DLL의 지터를 측정하고 시뮬레이션 결과로 얻어진 임피던스 값과 비교하였다. PDN의 공진 피크가 100MHz 주파수에서 1옴보다 큰 임피던스를 갖기 때문에 DLL의 지터는 주파수가 100MHz 근처에서 증가함을 보여준다. 타겟 임피던스를 줄이기 위한 방법인 디커플링 커패시터에 따른 칩과 보드의 임피던스 변화를 보였다. 따라서 전원 공급망 설계는 디커플링 커패시터와 함께 코어 스위칭 전류와 I/O 스위칭 전류를 같이 고려해야 한다.

• 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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• 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.

• 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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• 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.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• 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.

• 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.

• 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.