• Nuclear Engineering and Technology
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• v.44 no.8
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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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• v.20 no.1
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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.

• Proceedings of the KSME Conference
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• 2001.11b
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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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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• v.53 no.5
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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.

• Nuclear Engineering and Technology
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• v.31 no.3
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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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• v.11 no.6
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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.

• Journal of Energy Engineering
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• v.10 no.3
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• pp.294-298
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• 2001

A research for transmutation reactor is in progress to transmute high radioactive isotopes into low radioactive ones. In this study, thermal-hydraulic and structural analysis was performed to design liquid metal target system that would be used in subcritical transmutation reactor. Diffuse plate installation was considered to enhance cooling of window. And thermal-structural analysis of window was performed varying window thickness, beam power, and coolant flow rate to determine target system design valuers. It is ensured that maximum window temperature and stress would be acceptable in the design condition.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.3996-4001
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• 2023

The High-flux Advanced Neutron Application Reactor (HANARO) produces radioisotopes (RIs) (¹³¹I, ¹⁹²Ir, etc.) through neutron irradiation on various RI production targets. Among them, ¹⁷⁷Lu and ¹⁶⁶Ho are particularly promising owing to their theranostic characteristics that facilitate simultaneous diagnosis and treatment. Prior to neutron irradiation, evaluating the nuclear heating of the RI production target is essential for ensuring the thermal-hydraulic safety of HANARO. In this study, the feasibility of producing ¹⁷⁷Lu and ¹⁶⁶Ho using irradiation holes of HANARO was investigated in terms of thermal-hydraulic safety. The nuclear heating rates of the RI production target by prompt and delayed radiation were calculated using MCNP6. The calculated nuclear heating rates were used as an input parameter in COMSOL Multiphysics to obtain the temperature distribution in an irradiation hole. The degree of temperature increase of the ¹⁷⁷Lu and ¹⁶⁶Ho production targets satisfied the safety criteria of HANARO. The nuclear heating rates and temperature distribution obtained through the in silico study are expected to provide valuable insight into the production of ¹⁷⁷Lu and ¹⁶⁶Ho using HANARO.

• Journal of the Korean Geotechnical Society
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• v.37 no.12
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• pp.25-31
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• 2021

Boundary conditions for thermal-hydraulic problems of soils play an essential role in the numerical accuracy. This study presents a boundary condition considering the thermo-hydraulic interaction between the ground and the atmosphere. Ground surface energy balance consists of solar radiation, ground radiation, wind convection, latent heat from water evaporation, and heat conduction to the ground. Equations for each heat flux are presented, and numerical analyses are performed in conjunction with the FEM program for the thermal-hydraulic phenomenon of unsaturated soils. Numerical results using the weather data at the Ulsan Meteorological Observatory are similar to the measured surface temperature. Latent heat caused by water evaporation during the daytime lowers the surface temperature of the bare soil, and a thermal equilibrium is reached at nighttime when the effect of the ground condition is significantly reduced. The temperature change of the surface ground is diminished at the deeper ground due to its thermal diffusion. Numerical analysis where the surface ground temperature is the primary concern requires considering the thermo-hydraulic interaction between the ground and the atmosphere.