• Journal of Radiation Protection and Research
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• 제35권4호
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• pp.172-177
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• 2010

중수로원전에서 환경으로 배출되는 방사성유출물의 양은 경수로원전에 비해 상대적으로 많고, 방사성유출물을 계속적으로 배출하는 연속배출(Continuous release) 방식으로 운용되고 있다. 이 때문에 원자로건물 배기 굴뚝(Stack) 등 주요 배출지점에 방사선검출기(Radiation detector)를 설치하여 방사성유출물의 농도를 실시간으로 감시하고 있다. 또한 방사성핵종 별로 연간 배출 가능한 유도배출한계(Derived Release Limits: DRLs)를 정하고, 이들 설정 값을 초과하지 않도록 엄격하게 관리하고 있다. 본 논문은 중수로원전 방사성유출물에 대한 배출관리 방식, 유도배출한계의 설정기준, 설정 방법론과 설정 현황을 조사하여 검토하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.233-236
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• 2002

The high-pressurized spray nozzle is used f3r special washing and cutting with strong impact force. The performance of this nozzle, which focused on spray penetration and radial dispersion, was mainly investigated to maximize the momentum and minimize the flow loss. Hence, our experimental research was conducted by changing the aspect ratio ranging from 0 to 3 with nozzle outlet of 1.1. The spray trajectory far high-pressurized water was experimentally investigated using PDPA diagnostics, which was available at spray downstream region. As the spray at upstream near the nozzle exit did not show the improved disintegration. The results showed empirical correlation with regard to non-dimensional axial velocity distribution, spray penetration, and radial spreading rate with photographic visualization.

• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2001년도 제33회 춘계학술대회 개최
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• pp.96-99
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• 2001

The fatigue life of support bearings is one of the most critical factors for the performance of a control rod driving mechanism. They are operated at high temperature and high pressure and especially lubricated with dramatically low viscosity water. The support bearing is made of standardized 440C stainless steel, and it supports thrust load including the weight of the driving system and external force. Friction and wear characteristics of this material operating under severe lubrication condition is not well known yet, although it is expected to be changed with respect to temperature and boundary pressure. So the friction characteristics are investigated in sliding conditions using the reciprocating tribometer which can simulate the operating conditions. Highly purified water is used as lubricant, and the water is heated up and pressurized. Friction farce on the reciprocating specimens is monitored by the load cells. The results of the experiments are presented in this paper.

• 대한기계학회논문집B
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• 제28권3호
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• pp.247-255
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• 2004

The cooling characteristic of water-air mixed spray for high water mass flux is not well defined, compared to that of highly pressurized spray. A series of research program was planned to develop the boiling correlation for whole temperature range in case of water-air mixed spray with high water mass flux. The cooling experiments of hot steel surface with initial temperature of 820$^{\circ}C$ were conducted in unsteady state with relatively high water mass flux. A computer program was developed to calculate the heat flux inversely from measured data by three inserted thermocouples. Finally the effects of water and air mass flux on the averaged film boiling heat flux and wetting temperature were studied. In this 1st report, it is found that the boiling curve was similar to that of highly pressurized spray and the decreased slope of heat flux in film boiling region with respect to surface temperature became steep by increasing air mass flux. Also it is shown that, by increasing air mass flux, the averaged heat flux in film boiling region was increased, and then saturated and the wetting temperature was increased, and then decreased. Finally when the heat flux in film boiling region is compared with that of highly pressurized spray, it is known that the cooling is improved by introducing air up to 60%.

• Journal of Radiation Protection and Research
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• 제10권1호
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• pp.41-49
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• 1985

1,300 MWe 가압경수로 공동내에서 중성자의 흐름해석이 수행되었다. 중성자의 흐름을 해석하는데는 1차원 수송코드인 ANISN, 2차원 수송코드인 DOT3.5, 3차원 Monte Carlo 코드인 TRIPOLI-02와 이들을 접속시켜주는 DOTTRI 등의 전산코드가 이용되었고, 본 계산에 사용된 전산기는 IBM 3033형이었다. 계산된 선속 및 선량율은 900 MW 가압경수로의 공동내에서 측정한 측정치와 비교검토 되었고, 그 결과 중성자 군별로 약간의 오차는 있었으나 전체적으로 큰 오차는 없었다. 이 결과는 대용량의 원자로 차폐설계, 원자로보수시, 기타 원자로 공동내에 출입할 경우에 방사선방어상 필요한 방어수단을 제공하는데 기여하였다.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1126-1139
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• 2016

In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor ($K_{eff}$) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor.

• 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.

• 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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• 제49권1호
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• pp.134-140
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• 2017

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2005년도 추계학술대회논문집
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• pp.267-270
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• 2005

Finite element model of pressurized water reactor internals were obtained using ANSYS software package to analyze dynamic characteristics. The pressure vessel, hold-down ring, alinement key, core support barrel(CSB), upper guide structure(UGS) and fluid gap were fully modeled using structural solid element(SOLID45) and fluid element(FLUID80) which is one of element types. Also modal analysis using the above finite element model has been performed. As a result, it was found that the fundamental beam mode natural frequency of the CSB were 8.2 Hz, the shell mode one 14.5 Hz. To verify the Finite Element Analysis(FEA), we compare the analysis result with experimental data that is obtained from the plant IVMS(internal Vibration Monitoring System). The experimental results are good agreement with the FEA model.