• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.58-63
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• 1997

This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore defector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.186-195
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• 1997

When asymmetric thermal-hydraulic conditions occur between cold legs, the core inlet temperature will be nonuniform if the coolant is not mixed perfectly in the lower plenum. These uneven core inlet conditions may induce the change in core power distribution. Thus realistic prediction of thermal mixing is important in such abnormal conditions. In this study, reactor internals, which are scaled down as to conserve the flow area ratio, are set up in the model of KORI Unit 1 with the scaling factor of 1/710 by volume and coolant temperatures are measured beneath the lower core plate. Based on experimental results, the ability of COMMIX-1B code to simulate the coolant mixing phenomena in the lower plenum is estimated. The results show that complete mixing never occurs in any conditions and the mixing pattern is characterized according to the plant type.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4583-4590
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• 2023

This study aims at modeling the beryllium reflector poisoning under neutron irradiation conditions and calculating the impact of beryllium poisoning on the core parameters of ETRR-2 research reactor. The CITVAP code was used to calculate the neutron flux and parameters of ETRR-2 core with beryllium reflector elements. The neutron flux in each reflector element was calculated to solve the modeling equations for the atomic densities of lithium-6 (⁶Li), tritium-3 (³H), and helium-3 (³He) using the BERYL program. The results are discussed based on CITVAP calculations of the core excess reactivity and cycle length Full Power Days (FPD). Possible solutions to minimize the degradation due to beryllium poisoning are also discussed and compared based on calculations.

• Proceedings of the KIEE Conference
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• 2006.07b
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• pp.875-876
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• 2006

Superconducting fault currents(SFCLs) are expected to improve not only reliability but also stability of power systems. The analysis on current limiting operations of the flux-lock type SFCL, which consists of a flux-lock reactor wound an iron core and a YBCO thin film, was compared the open-loop with the closed-loop iron core of the subtractive polarity winding. In the SFCL, operation characteristics could be controlled by adjusting the inductances and the winding directions of the coils, then magnetic field induced in the iron core. The current limiting characteristics under the same experimental conditions were generated regardless of the iron core conditions. We confirmed that capacity of the SFCL was increased effectively by the closed-loop iron core. However, the power burden of the system could be lowered by the open-loop iron core.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2048-2054
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• 2022

The coolant temperature feedback coefficient is an important parameter of reactor core power control system. To study the coolant temperature feedback coefficient influence on the core power control system of small PWR, the core power control system is built with the nonlinear model and fuzzy control theory. Then, the uncertainty quantification method of reactor core parameters is established based on the Latin hypercube sampling method and the Bootstrap method. Finally, under the conditions of reactivity step perturbation and coolant inlet temperature step perturbation, uncertainty analysis for two cases is carried out. The result shows that with fuzzy controller and fuzzy PID controller, the uncertainty of the coolant temperature feedback coefficient affects the core power control system, and the maximum uncertainties of core relative power, coolant temperature deviation, fuel temperature deviation and total reactivity are acceptable.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.933-940
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• 2024

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

• Proceedings of the KIPE Conference
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• 1999.07a
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• pp.65-70
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• 1999

In this paper, the cost-effective design of output reactor which is used to suppress the over-voltage at the motor terminal in the Adjustable Speed Drives(ASD) application is proposed. In the elevator drive system, the power cable length is relatively shorter than other ASD applications and then the over-voltage at the motor terminal depends on the frequency characteristics of the output reactor at the over-voltage operating frequency. The over-voltage suppression mechanism of output reactor in ASD application is analyzed and the dominant parameters of output reactor for the over-voltage suppression are extracted. Using these parameters as the design values and considering the high frequency characteristics of iron core in the reactor, a new cost-effective structure of output reactor is proposed. Experimental results of the conventional reactor and the proposed reactor with a 15kW induction motor are given to verify the proposed scheme.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.61-61
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• 1999