• 대한기계학회논문집B
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• 제30권9호
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• pp.825-833
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• 2006

To investigate the moderator coolability for CANDU-6 reactors, a test facility (HU-KINS) has been manufactured as a 1/8 scaled-down of a calandria tank. In the design of the test facility, a scaling law was developed in such a way to consider the thermal-hydraulic characteristics of a CANDU-6 moderator. The proposed scaling law takes into consideration of the energy conservation, the dynamic similitude such as dimensionless numbers, Archimedes number (Ar) and Reynolds number (Re), and thermal-hydraulic properties similitude. Using this proposed scaling law, the thermal-hydraulic scaling analyses of similar test facilities such as the SPEL (1/10 scale) and the STERN (1/4 scale), have been identified. As a result, in the case of the SPEL, while the energy conservation is well defined, the similarities of Ar and the heat density are not well considered. As for the similarity of the STERN, while both the energy conservation and the characteristics of Ar are well defined, the heat density is not. In the meanwhile, the HU-KINS test facility with 1/8 length scaled-down is well similitude in compliance with all similarities of the energy conservation, the fluid dynamics and thermal-hydraulic properties. To verify the adequacy of the similarities in terms of thermal-hydraulics, a computational fluid dynamic (CFD) analysis has been conducted using the CFX-5 code. As the results of the CFD analyses, the predicted flow patterns and variation of axial properties inside the calandria tank are well consistant with those of previous studies performed with FLUENT and this implies that the present scaling method is acceptable.

• Advances in Energy Research
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• 제5권2호
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• Tribology and Lubricants
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• 제20권1호
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• pp.33-40
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• 2004

This study represents the newly advanced formulation of hydraulic fluids for extended drain interval and introduces the performance results of used oil samples from various excavators. The used oil samples, in this paper, show that there is a sharp change in viscosity drop and moderate additive depletion when viscosity index of hydraulic oil is very high. For the extension of hydraulic fluid life, it is necessary to improve the stability of viscosity and oxidation. New target properties from the used oil analysis were proposed for extended life. Performance of newly developed hydraulic oil based on used oil analysis is compared with previously used one. The properties of new formulation are the viscosity index of 140 and improved thermal stability consists of VHVI base oil. Field test results showed the possibility of extension of fluid life. Additionally, for development of high performance product, new required propertied and performances were discussed.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1556-1568
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.3-8
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• 2001

For the successful design of nuclear reactor, it is very important to investigate thermal-hydraulic characteristics of fuel rod bundle. Fluid flow and heat transfer in the non-circular cross-section of nuclear fuel rod bundle are different from those found in common circular tube. And complex three dimensional flow including secondary and vortex flow, is formed around the bundles. The purpose of this research is to examine how geometries and flow conditions affect heat transfer in fuel rod bundle. Design data for nuclear fuel rod bundle and structure are surveyed, and $3{\times}3$ sub-channel model is adopted in this study. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.639-644
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• 1997

The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART(System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary phenomena Identification and Ranking Table(PIRT) has been developed based on the experts' knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP(Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.344-363
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• 1999

A multi-dimensional thermal-hydraulic system analysis code, MARS 1.3.1, has been developed in order to have the realistic analysis capability of two-phase thermal-hydraulic transients for pressurized water reactor (PWR) plants. As the backbones for the MARS code, the RELAP5/MOD3.2.1.2 and COBRA-TF codes were adopted in order to take advantages of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the MARS code, all the functional modules of the two codes were unified into a single code first. Then, the source codes were converted into the standard Fortran 90, and then they were restructured using a modular data structure based on "derived type variables" and a new "dynamic memory allocation" scheme. In addition, the Windows features were implemented to improve user friendliness. This paper presents the developmental work of the MARS version 1.3.1 including the hydrodynamic model unification, the heat structure coupling, the code restructuring and modernization, and their verifications.their verifications.

• Corrosion Science and Technology
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• 제11권6호
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• pp.218-224
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• 2012

Liquid droplet impingement erosion (LDIE) known to be generated in aircraft and turbine blades is recently appeared in nuclear piping. UT thickness measurements with both A-scan and B-scan UT inspection equipments were performed for a component estimated as susceptible to LDIE in feedwater heater vent system. The thickness data measured with B-Scan equipment were compared with those of A-Scan. Thermal hydraulic analysis based on ANSYS FLUENT code was performed to analyze the behavior of liquid droplets inside piping. The wall thinning rate and residual lifetime based on both existing Sanchez-Caldera equation and measuring data were also calculated to identify the applicability of the existing equation to the LDIE management of nuclear piping. Because Sanchez-Caldera equation do not consider the feature of magnetite formed inside piping, droplet size, colliding frequency, the development of new evaluation method urgently needs to manage the pipe wall thinning caused by LDIE.

• 에너지공학
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• 제10권3호
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• pp.294-298
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• 2001

사용 후 핵연료의 고독성 장수명 핵종을 저독성 단수명 핵종으로 변환시키기 위한 미임계 핵변환로 연구가 진행중이다. 본 논문에서는 이러한 미임계 핵변환로에서 사용될 표적 시스템을 설계하기 위하여 표적시스템에 대한 열적, 구조적 분석을 수행하였다. 표적시스템의 열수력 분석에서는 diffuse plate를 삽입함으로써 빔창의 냉각효과를 증대시킬 수 있었다. 또한, 주요 인자인 빔창두께, 빔출력, 냉각재 유량 변화에 따른 빔창의 열적, 구조적 건전성 분석을 수행하여 표적시스템의 설계치를 설정하였다. 본 설계조건 하에서 빔창의 최대 온도 및 음력은 허용가능한 범위에 있음을 확인하였다.