• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.799-806
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• 2012

In recent years, a new algorithm has been introduced to perform the regional overpower protection (ROP) detector layout optimization for $CANDU^{(R)}$ reactors. This algorithm is called DETPLASA. This algorithm has been shown to successfully come up with a detector layout which meets the target trip set point (TSP) value. Knowing that these ROP detectors are placed in a number of safety channels, one expects that there is an optimal placement of the candidate detectors into these channels. The objective of the present paper is to show that a slight improvement to the TSP value can be realized by optimizing the channelization of these ROP detectors. Depending on the size of the ROP system, based on numerical experiments performed in this study, the range of additional TSP improvement is from 0.16%FP (full power) to 0.56%FP.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.113-150
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• 2012

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

• Journal of computational fluids engineering
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• v.21 no.3
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• pp.89-97
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• 2016

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.2
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• pp.112-119
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• 2016

Many studies have been conducted on the performance of a passive autocatalytic recombiner (PAR), but not many have focused on the locations where the PAR is installed. During a severe accident in a nuclear reactor containment, a large amount of hydrogen gas can be produced and released into the containment, leading to hydrogen deflagration or a detonation. A PAR is a hydrogen mitigation method that is widely implemented in current and advanced light water reactors. Therefore, for this study, a PAR was installed at different locations in order to investigate the difference in hydrogen reduction rate. The results indicate that the hydrogen reduction rate of a PAR is proportional to the distance between the hydrogen induction location and the bottom wall.

• The magazine of the Korean Society for Advanced Composite Structures
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• v.5 no.1
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• pp.19-25
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• 2014

본 연구에서는 국내외 항공기 충돌에 대한 원자력발전소의 안전성 평가 및 규제 현황과 연구용원자로의 안전성 평가 및 규제 현황을 살펴보았다. 이러한 현황과 평가와 관련하여 연구용원자로에 적용할 수 있는 항공기 충돌 안전성 평가 및 대비설계 기준을 원자력발전소에 적용되는 기준을 기반으로 정리하였다. 본 연구를 바탕으로 후속되는 연구에서는 연구용원자로에 대한 실질적인 항공기 충돌 안전성 평가 및 대비설계 기준을 도출할 수 있으며, 평가 및 대비설계 방법을 상세히 정립할 수 있을 것이다. 결과적으로, 이를 바탕으로는 항공기 충돌에 대비한 연구용원자로 건물의 예비개념설계 모델을 개발할 수 있을 것으로 예상된다.

• Journal of Welding and Joining
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• v.19 no.1
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• pp.88-94
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• 2001

DUPIC (Direct use of spent PWR fuel in CANDU reactors) nuclear fuel is a CANDU fuel fabricated remotely from spent PWR fuel materials in a hot cell. The soundness of the end closure welds of nuclear fuel elements is an important factor for the safety and performance of nuclear fuel. To evaluate the soundness of the end closure welds of DUPIC fuel element, a precise X-ray inspection system is developed using a micro-focus X-ray generator with an image intensifier and a real time camera system. The fuel elements made of Zircaloy-4 and stainless steel by an Nd:YAG laser welding and a TIG welding aye inspected by the developed inspection system. The soundness of the welds of the fuel elements was confirmed by the X-ray inspection process, and the irradiation test of DUPIC fuel elements has been successfully completed at the HANARO research reactor.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.15 no.5 s.98
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• pp.612-619
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• 2005

In this work. two dynamic absorbers are introduced and designed to reduce the vibration of the large-size pressure vessel of a reactor for a petrochemical plant. The vibration modes and harmonic responses of the vessel are firstly analyzed by the finite element method. On the basis of the analyzed results, two dynamic absorbers are designed by a simple design theory. Furthermore, an optimization process is executed and an optimal design of the dynamic absorber is obtained to improve performance and structural safety of the vessel. As a result, the maximum displacement and stress of the vessel is decreased about $85\%$ and $65\%$ respectively, the design criteria being satisfied.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.636-643
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• 1995

The accurate determination of the fuel-cladding gap conductance as functions of rod burnup and power level may be a key to the design and safety analysis of a reactor. The incorporation of a sophisticated gap conductance model into nuclear design code for computing thermal hydraulic feedback effect has not been implemented mainly because of computational inefficiency due to complicated behavior of gap conductance. To avoid the time-consuming iteration scheme, simplification of the gap conductance model is done for the current design model. The simplified model considers only the heat conductance contribution to the gap conductance. The simplification is made possible by direct consideration of the gas conductivity depending on the composition of constituent gases in the gap and the fuel-cladding gap size from computer simulation of representative power histories. The simplified gap conductance model is applied to the various fuel power histories and the predicted gap conductances are found to agree well with the results of the design model.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.240-250
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• 1997

Liquid sodium is widely used as a coolant of LMR(Liquid Metal Reactor) because of its physical and nuclear properties. However, the liquid sodium is very chemically reactive with oxygen and water so that the study on the sodium fire plays an important role in the LMR safety analysis. In this study, a sodium fire model is suggested to analyze the sodium pool fire where both the flame and the reaction products are considered. And also, sodium pool fire analysis computer code, SOPA, is developed. The sensitivity study on the experimental parameters such as the thermal radiation from flame to atmospheric gas, the vessel cooling and the duration of sodium spill was performed. The results showed good agreements with experimental data in the literature.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.88-98
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• 2000

More than half of all incidents in large complex technological systems, particularly in nuclear power or aviation industries, were attributable in some way to human erroneous actions. These incidents were largely due to the human engineering deficiencies of man-machine interface (MMI). In nuclear industry, advanced computer-based MMI designs are emerging as part of new reactor designs. The impact of advanced MMI technology on the operator performance, and as a result, on plant safety should be thoroughly evaluated before such technology is actually adopted in nuclear power plants. This paper discusses the applicability of human reliability analysis (HRA) to support the design review process. Both the first-generation and the second-generation HRA methods are considered focusing on a couple of promising HRA methods, i.e., ATHEANA and CREAM, with the potential to assist the design review process.