• Nuclear Engineering and Technology
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• v.30 no.4
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.590-595
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• 1998

차세대 원자로(KNGR : Korea Next Generation Reactor)에는 새로운 안전개념으로서 피동형 안전주입탱크(Safety Injection Tank. SIT)의 도입을 고려하고 있는데, 이러한 피동형 유량조절기능은 안전주입탱크내의 유체기구(Fluidic device)인 볼텍스챔버(vortex chamber)에 의해 이루어진다. 볼텍스챔버는 내부에서 발생되는 와류강도에 따라 유동저항의 강도가 달라짐을 이용하여 유량을 피동적으로 조절할 수 있는 유체기구이다. 본 연구에서는 볼텍스챔버의 유동특성을 관찰하기 위하여 소규모 실험장치를 구축하고, 이를 이용하여 실험을 수행하였다. 본 연구는 두 단계로 수행되었다. 제1단계 실험에서는 볼텍스챔버의 기하학적 특성이 안전주입탱크의 안전주입수 방출특성에 미치는 영향에 대한 거시적 관점에서의 연구로서. 볼텍스챔버의 기하학적 변수(유입구 및 방출구의 직경)가 안전주입수의 방출과정에서 발생되는 SIT 내의수위 거동, 안전주입수의 방출유량 특성등에 미치는 영향에 대해 중점적으로 고찰하였다 제2단계 실험에서는 1단계 실험에서 관찰된 안전주입탱크의 여러 가지 방출특성과 볼텍스챔버 내부 유동장의 유동특성과의 관련성을 규명하기 위해 PIV (Particle Image Velocimetry)를 이용하여 볼텍스챔버의 기하학적 변수에 따른 유동장 내부의 국소 유속분포를 측정하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.599-604
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• 1997

One of important inconsistencies in the six-equation model predictions has been found to be the force experienced by a single bubble placed in a convergent stream of liquid. Various sets of governing equations yield different amount of forces to hold the bubble stationary in a convergent nozzle. By using the first order potential flow theory, it is found that the six-equation model can not be used to estimate the force experienced by a deformed bubble. The theoretical value of the particle stress of a bubble in a convergent nozzle flow has been found to be a function of the Weber number when bubble distortion is allowed. This force has been calculated by using different sets of governing equations and compared with the theoretical value. It is suggested in this study that the bubble size distribution function can be used to remove the presented inconsistency by relating the interfacial variables with different moments of the bubble size distribution function. This study also shows that the inconsistencies in the thermal-hydraulic governing equation can be removed by mechanistic modeling of the phasic interface.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.23-29
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• 2011

The renewed global interest in nuclear power has arisen from the need to reduce greenhouse gas emissions and to provide sufficient electricity for a growing global population before the accident at Fukushima Dai-ichi nuclear power plant in Japan. In spite of the safety issues of nuclear power plants raised by the ongoing Japanese nuclear crisis, many countries with nuclear power plants (NPPs) are still implementing license extensions of 10~20 years, and even consideration is being given to the concept of life-beyond-60, a further period of license extension from 60 to 80 years. To solving the materials aging problem is integral to its success. To evaluate the plant aging phenomena, a lot of background information such as materials and environment of the parts of the reactor and plant systems is needed by the experts. Information on degradation mechanisms is also used. In this paper, a materials degradation evaluation program called OnMDE-SYS (On-line Materials Degradation Evaluation System) is introduced. The developed program provides a variety of information on the materials and stressors as well as operational experience to the experts. It is also anticipated that the experts can perform materials degradation assessment on the web directly by referring to domestic and international information about the degradation of a nuclear power plants through OnMDE-SYS.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.311-318
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• 1992

During mid-loop operation of Nuclear Power Plant, to prevent the Decay Heat Removal System (DHRS) from failure due to air entrainment of free surface vortex in the piping system, a set of simulating experiments was performed. Through these experiments, a relation between the non-dimensionalized numbers, such as H/d, Froude number, Reynolds number, was found. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from viewpoint of reactor safety, a modified inlet device which is reducer type is strongly recommended for the prevention of air entrainment into DHRS.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2236-2241
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• 2008

The surface heat flux of nuclear fuel rod is the most important factor which can affect safety of reactor and fuel. If fuel rod surface heat flux exceeds the CHF(${\underline{C}}ritical$ ${\underline{H}}eat$ ${\underline{F}}lux$), fuel can be damaged. In case of double cooled annular fuel, which is under developing, contains two coolant channels. Therefore, a generated heat in the fuel pellet can move to inner or outer channel and heat flow direction is decided by both sides heat resistance which varied by dimension and material property change which caused by temperature and irradiation. The new program(called DUO) was developed. For the calculation of surface heat flux, a both sides convection by inner/outer coolant, s gap temperature jump and conduction in the fuel are modeled. Especially, temperature and time dependent fuel dimension and material property change are considered during the iteration. A sample calculation result shows that the DUO program has sufficient performance for annular fuel thermal hydraulics design.

• Journal of Magnetics
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• v.22 no.1
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• pp.34-42
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• 2017

Material defects in the Steam Generator Tubes (SGT) of sodium cooled fast breeder reactor (PFBR) can lead to leakage of water into sodium. The water and sodium reaction will lead to major accidents. Therefore, the examination of steam generator tubes for the early detection of defects is an important requirement for safety and economic considerations. In this work, the Magnetic Flux Leakage (MFL) based Non Destructive Testing (NDT) technique is used to perform the defect detection process. The rectangular notch defects on the outer surface of steam generator tubes are modeled using COMSOL multiphysics 4.3a software. The obtained MFL images are de-noised to improve the integrity of flaw related information. Grey Level Co-occurrence Matrix (GLCM) features are extracted from MFL images and taken as input parameter to train the neural network. A comparative study on characterization have been carried out using feed-forward back propagation (FFBP) and cascade-forward back propagation (CFBP) algorithms. The results of both algorithms are evaluated with Mean Square Error (MSE) as a prediction performance measure. The average percentage error for length, depth and width are also computed. The result shows that the feed-forward back propagation network model performs better in characterizing the defects.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.237-246
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• 1994

A high flow critical heat flux (CHF) correlation, based on the single-pin CHF experimental data for finned and unfinned heated rods, was developed for the thermal-hydraulic design and safety analysis of the Korea Multi-purpose Research Reactor (KMRR) core. The correlation consists of dimensionless parameters such as Reynolds number, thermodynamic equilibrium quality, liquid-to-vapor density ratio, and hydraulic equivalent diameter ratio. The fin effect was taken into account in the correlation by a finned-to-unfinned heated perimeter ratio. The effects of a cold wall and non-uniform axial power distribution ore discussed to verify the applicability of the single-pin based correlation to the KMRR fuel bundle. The correlation limit departure from nucleate boiling ratio (DNBR) was determined as 1.44 from the statistical analysis of the CHF data.

• Nuclear Engineering and Technology
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• v.40 no.5
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• pp.365-374
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• 2008

Dependability-critical systems, such as digital instrumentation and control systems in nuclear power plants, necessitate engineering techniques and tools to provide assurances of their safety and reliability. Determining system reliability at the architectural design phase is important since it may guide design decisions and provide crucial information for trade-off analysis and estimating system cost. Despite this, reliability and system engineering remain separate disciplines and engineering processes by which the dependability analysis results may not represent the designed system. In this article we provide an overview and application of our approach to build architecture-based, dynamic system models for dependability-critical systems and then automatically generate dynamic fault trees (DFT) for comprehensive, tool-supported reliability analysis. Specifically, we use the Architectural Analysis and Design Language (AADL) to model the structural, behavioral and failure aspects of the system in a composite architecture model. From the AADL model, we seek to derive the DFT(s) and use Galileo's automated reliability analyses to estimate system reliability. This approach alleviates the dependability engineering - systems engineering knowledge expertise gap, integrates the dependability and system engineering design and development processes and enables a more formal, automated and consistent DFT construction. We illustrate this work using an example based on a dynamic digital feed-water control system for a nuclear reactor.