• Proceedings of the KIEE Conference
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• 1986.07a
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• pp.429-432
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• 1986

• Nuclear Engineering and Technology
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• v.15 no.1
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• pp.50-56
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• 1983

The effect of flow stratification is of particular concern during transient after scram in the outlet plenum of LMFBR. In this case, buoyancy effects on turbulent mixing are of importance to designers. An investigation has been made to identify the appropriate change in the available turbulence models which are necessary to include the effects of buoyancy on turbulence transport equations. The developed physical model of the buoyant turbulent flow are solved through SMAC method. Testing of the developed numerical model was undertaken and compared with experimental results. The results show that the buoyant turbulent effects account for a significant increase in the stability of the stratification, with a strong suppression of turbulence in the outlet plenum.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.7
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• pp.1042-1049
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• 1997

LMFBR(Liquid Metal Fast Breeder Reactor) vessel is operated under the high temperatures of 500-550.deg. C. Thus, transient thermal loads were severe enough to cause inelastic deformation due to creep-fatigue and plasticity. For reduction of such inelastic deformations, Y-piece structure in the form of a thermal sleeve is used in LMFBR vessel under repeated start-up, service and shut-down conditions. Therefore, a systematic method for inelastic analysis is needed for design of the Y-piece structure subjected to such loading conditions. In the present investigation, finite element analysis of heat transfer and inelastic thermal stress were carried out for the Y-piece structure in LMFBR vessel under service conditions. For such analysis, ABAQUS program was employed based on the elasto-plastic and Chaboche viscoplastic constitutive equations. Based on numerical data obtained from the analysis, creep-fatigue damage estimation according to ASME Code Case N-47 was made and compared to each other. Finally, it was found out that the numerical predictio of damage level due to creep based on Chaboche unified viscoplastic constitutive equation was relatively better compared to elasto-plastic constitutive formulation.

• Nuclear Engineering and Technology
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• v.12 no.2
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• pp.88-98
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• 1980

The process of solidification of a single-phase flowing hot fluid in a cylindrical tube has been investigated analytically and experimentally. A series of tests were performed, using paraffin -wax and Wood's metal as flowing hot fluids. These data verified the existing quasistatic numerical analysis model of freezing process developed at Brookhaven National Laboratory In addition, experimental results provided information regarding the effects of various parameters on the .process of transient flowing and freezing through a vertical channel. The experimental apparatus and techniques are described. Comparison of experimental data with predictions of mathematical models for transient molten fluid displacement are presented in graphical form. In addition, the mathematical model is applied to LMFBR post-accident conditions.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.473.1-473
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• 2001

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.267-275
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• 1999

This paper deals with the industrial problem of thermal striping damage on the French prototype fast breeder reactor, Phenix and it was studied in coordination with the research program of IAEA. The thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the tee-junction of the secondary piping using Green's function method and standard FEM is presented. The thermohydraulic(T/H) loading condition used in the present analysis is the random type thermal loads computed by T/H analysis on the turbulent mixing of the two flows with different temperatures. The thermomechanical fatigue damage was evaluated according to ASME code section 111 subsection NH. The results of the fatigue analysis showed that fatigue failure would occur at the welded joint within 90,000 hours of operation. The assessment for the fracture behavior of the welded joint showed that the crack would be initiated at an early stage in the operation. It took 42,698.9 hours for the crack to propagate up to 5 mm along the thickness direction. After then, however, the instability analysis, using tearing modulus, showed that the crack would be arrested, which was in agreement with the actual observation of the crack. An efficient analysis procedure using Green's function approach for the crack propagation problem under random type load was proposed in this study. The analysis results showed good agreement with those of the practical observations.

• Proceedings of the KIEE Conference
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• 2003.07b
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• pp.906-908
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• 2003

A Linear induction electromagnetic(EM) pump of liquid metal fast breeder reactor(LMFBR) is used for the purpose that the liquid metal of high temperature is transported by EM force. This paper evaluates magnetic flux density necessary for transporting liquid metal, using analytical model of the linear induction EM pump. Using the 2-D finite element method(2-D FEM), magnetic flux density is estimated in consideration of a geometric model, electric parameter, and velocity of liquid metal. From the viewpoint of hydrodynamics, the results can be used for flow analysis of the liquid metal.

• Nuclear Engineering and Technology
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• v.4 no.2
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• pp.109-115
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• 1972

The evolution of radioiodine 131I from a sodium peroxide system as a function of time, temperature, and carrier gas (nitrogen) flow rate was studied. Virtually no iodine was volatilized at 25$0^{\circ}C$ and a very small amount, of the order of 10$^{-3}$ % per flour, at 63$0^{\circ}C$. Substantially greater amounts of iodine were volatilized at 7$25^{\circ}C$ and 83$0^{\circ}C$. The data are consistent with the hypothesis that the mechanism of transfer is distillation of sodium iodide, and that elemental iodine is not produced in this system.

• Nuclear Engineering and Technology
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• v.19 no.2
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• pp.85-98
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• 1987

To analyze the aerosol dynamics in severe accidents of LMFBR, a new computer code entitled MCAD (Multicomponent Aerosol Dynamics) has been developed. The code can treat two component aerosol system using relative collision probability of each particles as sequences of accident scenarios. Coagulation and removal mechanisms incorporating Brownian diffusion and gravitational sedimentation are included in this model. In order to see the effect of particle geometry, the code makes use of the concept of density correction factor and shape factors. The code is verified using the experimental result of NSPP-300 series and compared to other code. At present, it fits the result of experiment well and agrees to the existing code. The input variables included are very uncertain. Hence, it requires uncertainty and sensitivity analysis as a supplement to code development. In this analysis, 14 variables are selected to analyze. The input variables are compounded by experimental design method and Latin hypercube sampling. Their results are applied to Response surface method to see the degree of regression. The stepwise regression method gives an insight to which variables are significant as time elapse and their reasonable ranges. Using Monte Carlo Method to the regression model of LHS, the confidence level of the results of MCAD and their variables is improved.

• Nuclear Engineering and Technology
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• v.32 no.6
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• pp.566-594
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• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.