• 한국안전학회지
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• 제14권3호
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• pp.3-11
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• 1999

The atomic fuel rods between top and bottom end pieces of reactor need to be extended for high combustion rate of future-type fuel to increase the irradiation in the axial direction. For allowing axial extension of the fuel rods, the space between top and bottom end pieces should be expanded. Thus the thickness reduction of the flow plate is necessary. This study was carried out the mechanical strength test by using strain gages as a function of flow plate thickness, the existence of skirt and loading condition for the Korean Fuel Assembly(KOFA). The experimental apparatus was designed for load conditions, uniformly distributed load and displacement. Test method using whiffle tree of uniformly distributed load has been comparatively conservative. The test results were compared with those of finite element analysis and the test method on bottom end piece was established.

• 대한전자공학회:학술대회논문집
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• 대한전자공학회 2002년도 하계종합학술대회 논문집(5)
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• pp.153-156
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• 2002

The remotely operated underwater robotic vehicle system has been required to inspect some objects such as baffle former bolts and remove foreign objects in reactor vessel of nuclear power plant. In this paper, we have designed the remotely operated underwater robotic vehicle system that includes a long reach arm that is composed of 4 joints to remove foreign objects in a narrow space, a camera for visual test, instrument sensors for vehicle positioning, 4 thrusters for underwater navigation of vehicle, and supervisory control system implemented with industrial PC that includes robot simulator that has the functions of real time visualization, robot work planning and etc.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2008년도 춘계학술대회논문집
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• pp.641-646
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• 2008

Long and slender body with or without flexible supports under severe operating condition can be unstabilized even by the small cross flow. Turbulent flow mixer, which actually increases thermal-hydraulic performance of the nuclear fuel by boosting turbulence, disturbs the flow field around the fuel rod and affects dynamic behavior of the nuclear fuel rods. Few studies on this problem can be found in the literature because these effects depend on the specific natures of the support and the design of the system. This work shows how the dynamics of a multi-span fuel rod can be affected by the turbulent flow, which is discretely activated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was established. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• 센서학회지
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• 제25권4호
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• pp.264-270
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• 2016

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1135-1142
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• 2017

The applicability of the algebraic collapsing acceleration (ACA) technique to the method of characteristics (MOC) in cases with scattering anisotropy and/or linear sources was investigated. Previously, the ACA was proven successful in cases with isotropic scattering and uniform (step) sources. A presentation is first made of the MOC implementation, available in the DRAGON5 code. Two categories of schemes are available for integrating the propagation equations: (1) the first category is based on exact integration and leads to the classical step characteristics (SC) and linear discontinuous characteristics (LDC) schemes and (2) the second category leads to diamond differencing schemes of various orders in space. The acceleration of these MOC schemes using a combination of the generalized minimal residual [GMRES(m)] method preconditioned with the ACA technique was focused on. Numerical results are provided for a two-dimensional (2D) eight-symmetry pressurized water reactor (PWR) assembly mockup in the context of the DRAGON5 code.

• 센서학회지
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• 제25권3호
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• pp.229-234
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• 2016

Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the vanadium (V) SPND has been being developed to be used in OPR1000 nuclear power plants. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina insulator with a cylindrical geometry. An MCNP code was used to simulate some factors (neutron self-shielding factor and beta escape probability from the emitter) and space charge effect of an insulator necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND and contribute to the development of TMI (Top-mount In-core Instrumentation) which might be used in the SMART and SMR.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• Nuclear Engineering and Technology
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• 제38권6호
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• pp.459-482
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• 2006

A geochemical approach to the long-term safety of waste disposal is discussed in connection with the significance of actinides, which shall deliver the major radioactivity inventory subsequent to the relatively short-term decay of fission products. Every power reactor generates transuranic (TRU) elements: plutonium and minor actinides (Np, Am, Cm), which consist chiefly of long-lived nuclides emitting alpha radiation. The amount of TRU actinides generated in a fuel life period is found to be relatively small (about 1 wt% or less in spent fuel) but their radioactivity persists many hundred thousands years. Geological confinement of waste containing TRU actinides demands, as a result, fundamental knowledge on the geochemical behavior of actinides in the repository environment for a long period of time. Appraisal of the scientific progress in this subject area is the main objective of the present paper. Following the introductory discussion on natural radioactivities, the nuclear fuel cycle is briefly brought up with reference to actinide generation and waste disposal. As the long-term disposal safety concerns inevitably with actinides, the significance of the aquatic actinide chemistry is summarized in two parts: the fundamental properties relevant to their aquatic behavior and the geochemical reactions in nanoscopic scale. The constrained space of writing allows discussion on some examples only, for which topics of the primary concern are selected, e.g. apparent solubility and colloid generation, colloid-facilitated migration, notable speciation of such processes, etc. Discussion is summed up to end with how to make a geochemical approach available for the long-term disposal safety of nuclear waste or for the performance assessment (PA) as known generally.

• Nuclear Engineering and Technology
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• 제18권2호
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• pp.71-77
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• 1986

원자로 노심을 축방향으로 일차원 해석을 하고, 가입경수로형원자로의 안전성 해석에 적용할 수 있는 중성자 동특성프로그램 BIK를 개발하였다. BIK프로그램내에서 공간변수에 대해서는 유한차분법이, 시간변수에 대해서는 $\theta$-시간적분법이 채택되었다. 또한 도플러 및 감속재 궤환과 제어봉구동 등을 자세히 묘사하는 모델들이 포함되었다 핵모델의 검증은 ANL검증문제를 통해 이루어졌고, 고리 1호기의 제어봉 인출사고시의 노심출력 변화를 계산하였다. 이상의 계산결과 BIK동특성프로그램이 노심의 중성자 속 변화를 일차원해석의 한계내에서 비교적 정착하게 묘사할 수 있으며, 가압경수로형 원자로의 안전성 해석에 유용하게 사용될 수 있다는 것이 증명되었다.

• 대한산업공학회지
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• 제26권1호
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• pp.54-65
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• 2000

The objective of this study is to evaluate the operator performance relating to hand switches with two or three buttons in the main control rooms of nuclear power plants. Based on the comparative analysis of the nuclear power plants, two different subjective workload-rating scales were used to evaluate the performance of 48 operators: the Overall Workload(OW) and National Aeronautics and Space Administration Task Load Index (NASA TLX). The survey questions consisting of the eight-items were asked to evaluate the operating experiences for the two different switch types. The OW scales ratings were applied to measure the workload of the switch-related tasks. The ratings revealed that signal detection tasks caused less workload in the three-buttoned-switch operators than the other switch group. However, in the switch operation tasks, the switch types did not show statistically significant effects on workload level. The NASA TLX scale ratings were performed based on detailed task scenarios that assumed the accident of small break loss of coolant, what we call, the small LOCH. The NASA TLX was administered to three different task groups: the reactor, the turbine, and the electric operator groups. Based on the NASA TLX, the two-buttoned switch groups showed higher workload than those with the three-buttoned switches. However, a statistically significant difference was found only in the reactor operator groups. When the current switch type was assumed to be changed for the other type, all of the three-buttoned switch groups were predicted to have higher workload than the other switch groups, respectively. The implications of these findings were discussed.